ML20081J709
| ML20081J709 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 10/31/1983 |
| From: | Cherdack R, Gahan E, Gallup D, Skala G SANDIA NATIONAL LABORATORIES |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| CON-FIN-A-1309 NUREG-CR-3421, NUDOCS 8311090088 | |
| Download: ML20081J709 (165) | |
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i NUREG/CR-3421 F AMQ83- 1629 AN Printed August 1983 l
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l, 4 i Cost / Benefit Analysis of Adding a Feed and Bleed Capability to Combustion Engineering Pressurized Water Reactors Donald R. Gallup, E. Gahan R. Cherdack, G. Skala Prepared by Sarxia National Laboratones >
Albuquerque, New Mexico 87185 and Lvermore. Cakforrua 94550 for the United States Department of Energy under Contra:t DE-AC04-76DP00789 I
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any legal liability or responsibility for any third party's use, or the results of such use, of any in'ormation, apparatus product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights.
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u COST / BENEFIT ANALYSIS OF ADDING A FEED AND BLEED CAPABILITY,TO COMBUSTION ENGINEERING PRESSURIZED WATER REACTORS 1
1 Donald R. Gallup Sandia National Laboratories E. Gahan R. Cherdack G. Skala Burns & Roe, Inc.
{ August 1983 Sandia National Laboratories-l Albuquerque, New Mexico 87185.
- operated by Sandia Corporation
. for the i U.S. . Depar tment of Energy l-Prepared for
, Division of Safety Technology
- Office of Nuclear.Raactor Regulation gr-) U.S. Nuclear Regulatory Commission Washington,. DC 20555 e Under Memorandum of Understanding DOE.40-550-75 t
NRC' Fin No. A1309 I
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l ABSTRACT 4
This report presents the results of a cost / benefit analysis 0- for the addition of a feed and bleed capability to the San Onofre o Nuclear Generating Station, Unit 2, (SONGS 2). Two cases of feed and bleed capability were investigated: 1) adding power operated relief valves (PORVs) to the pressurizer for depressurization and using the present high pressure safety injection (HPSI) system for reactor coolant system (RCS) inventory make-up and 2) adding an independent single train feed and bleed system. For the first case, it is estimated that the core melt frequency would be incrementally reduced by 4.0E-6 per year, a factor of 1.3, at a cost of $2.5 M to $4.3 M depending on when the equipment is installed. For the second case, it is estimated that the core melt frequency would be incrementally reduced by 1.2E-5 per year, a factor of 3, at a cost of $7.0 M to $10.3 M.
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CONTENTS Page e 1.0 Introduction ............................................ 1 2.0 Feedwater Reliability ................................... 3 2.1 MFWS and AFWS Reliability .......................... 3 2.2 Common Mode Failure ................................ 3 2.3 AFWS Hidden Failures ............................... 6 2.4 Conclusions on Feedwater Reliability ............... 8 3.0 PORV LOCAs .............................................. 9 3.1 Frequency of PORV LOCAs ............................ 9 3.2 Effects of PORV LOCAs .............................. 9 3.3 Frequency of a PORV LOCA at San Onofre 2 ........... 9 3.4 Conclusions on PORV Related LOCAs .................. 13 4.0 Benefits Obtained by Adding a Feed and Bleed Capability to San Onofre 2 .............................. 18 4.1 Reliability of the San Onofre 2 AFWS ............... 18 4.2 Event Trees ........................................ 18 4.3 Analysis Results ................................... 31 5.0 Costs Associated with Installing a Feed and Bleed Capability .............................................. 36 5.1 Study Methods ...................................... 36 5.2 Conceptual Design for Plant Upgrading .............. 38 5.3 Detailed Design, Installation and Oparation of a Feed and Bleed Capability ..................... 40 5.4 Implementation Cost Consideration .................. 41 5.5 Conclusions on Feasibility and Costs ............... 43 6.0 Cost-Benefit Analysis ................................... 45 References ................................................... 47 Appendix A. A Study of Controlled Depressurization of the Primary System in Pressurized Water Reactors .... A-1 Appendix B. Major Details of the Reliability Assessment O
Done for SONGS 2 ................................ B-1 t
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LIST OF FIGURES s Page p Figure 2.1: Simplified AFWS in Stand-by Mode ................. 7
- Figure 3.1: Fault Tree for Determining the Frequency of a PORV LOCA at San Onofre 2 ...............,....... 14 Figure 4.1: Simplified Drawing of the SONGS 2 AFWS ........... 19 Figure 4.2: Simplified Drawing of the SONGS 2 HPSI System .... 22 Figure 4.3: Simplified Drawing of a Potential Feed and Bleed System for SONGS 2 ........................ 23 Figure 4.4: Transient Event Tree for San Onofre 2 Without PORVs ................................... 24 Figure 4.5: SGTR - Small LOCA Event Tree for San Onofre 2 Without PORVs ............,...................... 26 Figure 4.6: Transient Event Tree for San Onofre 2 With PORVs ........................................... 27 Figure 4.7: SGTR - Small LOCA Event Tree for San Onofre 2 With PORVs ...................................... 28 Figure 4.8: Transient Event Tree for San Onofre 2 With a Dedicated Feed and Bleed System ................. 29 Figure 4.9: SGTR - Small LOCA Event Tree for San Onofre 2 With a Dedicated Feed and Bleed System ........... 30 r
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l LIST OF TABLES Page
. Table 2.1. MFWS and AFWS Reliability Data .................... 4 Table 2.2. Nomenclature Used in Table 2.1 .................... 5 Table 3.1. Data on PORV Initiated LOCAs ...................... 10 Table 3.2. Probability that a PORV LOCA which Leads to Core Melt will Result in a Certain Size Release ... 11 Table 3.3. Release Categories Used in the Reactor Safety Study ............................................ 12 Table 3.4. Data Used to Determine PORV LOCA Frequency ........ 17 Table 4.1. Reliability of the San Onofre 2 AFWS .............. 20 Table 4.2. Nomenclature, Assumptions, and Data Used in the San Onofre Event Trees ........................ 25 Table 4.3. Core Melt Frequencies for San Onofre 2 Without PORVs (Case 1) .................................... 32 Table 4.4. Core Melt Frequencies for San Onofre 2 With PORVs (Case 2) .................................... 33 Table 4.5. Core Melt Frequencies for San Onofre 2 with a Dedicated Feed and Bleed System (Case 3) .......... 34 Table 4.6. Comparison of Core Melt Frequencies at San Onofre With and Without a Feed and Bleed Capability ..... 35 Table 6.1. Cost / Benefit Comparison of Adding a Feed and Bleed Capability to the San Onofre 2 Nuclear Power Plant ....................................... 46 l
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1.0 INTRODUCTION
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Sandia National Laboratories has been requested, as a part of Task Action Plan A-45 (TAP A-45) on Decay Heat Removal, to evaluate o the costs and benefits of adding a rapid depressurization capabil-ity to Combustion Engineering ( C E .' nuclear power plants without power operated relief valves (PORVs). This request was made by Mr. Harold Denton, Director of NRC's office of Nuclear Reactor Regulation (NRR), and stems from the July 28, 1982, Commission Meeting on a Full Power Operating License for San Onofre Nuclear Generating Station (SONGS) 2, a CE-design, pressurized water reactor (PWR).
The concern about the lack of a rapid depressurization cap-ability on those CE plants without PORVs is that there may not be sufficient redundancy in the plant safety systems to provide adequate assurance that the reac tor can be operated safely.
Specifically, all transients, steam generator tube ruptures (SGTR) and some small loss of coolant accidents (LOCAs) on these CE plants require decay heat removal from the primary coolant system by the secondary coolant system. The only systems available to provide this function are the main feedwater system (MFWS) and auxiliary feedwater system (AFWS), and many transients (about 15%) involve che loss of function of the MFWS. By adding PORVs and consequently the ability to depressurize rapidly, it may be possible to remove decay heat from the reactor core by a process known as " feed and bleed". (In this process, steam and pressure are relieved from the primary system through PORVs, and reactor coolant inventory make-up is provided using high head injection pumps.) A feed and bleed system would provide a backup heat removal capability to the AFWS and MFWS in the event of a transient, SGTR or small LOCA.
A drawback to adding PORVs to these CE plants is that they provide a potential LOCA path. PORVs are normally designed to open automatically for pressure relief if the pressure of the reactor coolant system (RCS) reaches a certain point during a transient. If a PORV fails to reseat following such a transient and the relief of steam from the RCS cannot be isolated using other valves, a LOCA requiring plant safety system response will ensue.
The approach taken in this study to determine the change in core melt probability and associated costs as a result of adding PORVs to a CE plant is as follows. First, probabilistic risk assessments (PRAs) and other literature were reviewed to determine the reliability of the MFWS and AFWS at specific PWRs. Second, existing PRAs were used to determine the frequency and consequences of PORV-initiated LOCAs at the same power plants. These first two steps are a generalized response to questions 8 (a)-8 (d) and 10 of the NRC letter dated March 26, 1982, to Mr. A. E. Scherer, Director
, of Nuclear Licensing at Combustion Engineering.(1) They provide l general insight iato the reliability of AFWSs and MFWSs, and hence the need for feed and bleed systems, and also into the frequency and consequences of accidents that may be aggravated by equipment
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required for feed and bleed. systems. Third, the net benefit of installing PORVs at the San Onofre 2 power plant was determined for the cases when 1) the present high pressure safety injection -
(HPSI) system is used for reactor coolant make-up and 2) a single dedicated train of injection capability is installed for reactor coolant inventory make-up. (The analysis done for a feed and bleed train was not requested by Mr. Denton, but is commensurate with the broader objectives of TAP A-45.) Fourth, the cost of adding each of the two feed and bleed systems described in the third step was addressed. These latter two steps address questions 11 and 12 of the March 26 letter to Mr. Scherer. Finally, a cost /
benefit analysis was performed for adding a feed and bleed system to SONGS 2 using the results of the third and fourth steps.
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2.0 FEEDWATER RELIABILITY As discussed in the previous section, the issue of adding a feed and bleed capability to those CE plants which do not present-
- 'ly have that capability stems from a concern that the feedwater systems, both main and auxiliary, may not have sufficient reli-ability. It is essential in those plants without a feed and bleed capability, that feedwater be provided to the steam generators to remove decay heat from the reactor core in the event of a tran-sient, SGTR or some small LOCAs (i.e., LOCAs with a break size less than about 3 square inches). The approach taken in this study to address the question of feedwater reliability was to survey PRAs and other literature to obtain general insights into MFWS and AFWS reliability. Also, two special issues regarding feedwater reliability were addressed: 1) initiating events which could impact both the MFWS and AFWS reliabilities and 2) hidden single failures in MFWSs and AFWSs which could result in their failure.
2.1 MFWS and AFWS Reliability The reliability data obtained from PRAs on MFWSs and AFWSs are presented in Table 2.1. This table shows that the reliability of MFWSs does not var probabilityis10ggreatlyfromplanttoplant. The failure for most plants which took credit for MFWS operation given that the initiating event did not cause MFWS failure. (It should be noted that the value 10-2 was initially used in the Reactor Safety Study and used subsequently in the other PRAs).
2 The reliability of AFWSs, on the other hand, can vary significantly from plant to plant depending on the initiating event. For example, given loss of offsite power (LOSP) as the initiating event, there is a factor of about 1000 in the difference between the AFWS reliabilities for Zion 1 and Calvert Cliffs 2.
However, when all AC power is lost, the reliabilities of the AFWSs surveyed do not vary by more than a factor of 5. The reason for the large variance in the AFWS reliabilities lies in systeu design. The two design features which are the most significant contributors to this variance are: 1) the number of trains in the system and 2) the type of actuation utilized, manual or automa-tic. (It should be noted that modifications have been made to the AFWSs of some of the referenced plants'since the time of the' analyses.) Consequently, a plant-specific analysis must be done when addressing AFWS reliability issues. Such an analysis was-
, done for San Onofre 2, and the results are reported in Section 4.
2.2 Common Mode Failure l u The only common mode failure for AFWSs and MFWSs identified in the PRAs surveyed-was loss of DC power. Loss of DC' power is an initiating event which causes the loss of all plant instrumentation and control systems. The frequency of occurrence of loss of DC Table 2.1. MrWS and AFWS Reliability Datata)(b)
AFWS Q(L/T Q (L/T Q (L/ AC)
' Plant Pumps f(T 3)(h) f(7 2) Q(M) -R(M) Q(L/Tg ) Boundh) Q(L/Ty) Bound $) Q (L/AC) Bounds Surry 2 motor. .16/yr 3/yr .01(c) .01(8I 4.5E-4 2.8E-3 1 turbine .2(di 1.2E-4 3.8E-5 2.2E-5 2E-2 San Onofre 2 motor 2& 3 1 turbine
^ Crystal 1 motor .35/yr 1.8/yr .1 1.8E-3 3.4E-4 River 3 1 turbine Calvert 2 turbine .18/yr. 3/yr .01 (f ) .1(C) 3.9E-3 3E-3 Clitts 2 Sequoyah 1 2 motor .01(f3 4.3E-5 4.3E-5 1.9E-2 1 turbine II93 1.1E-5 4.7E-5 9.7E-3 3.6E-2
"'Inaian 2 motor .14/yr 6.7/yr 2.8E-6 3.9E-3 Point 2 1 turbine 3.8/yr 1(9) 7.6E-6 8.2E-5 1.2E-2 3.1E-2 Inalan 2 motor .14/yr 1.1E-6 5.4E-3 Point 3 1 turbine I. .1(c) 2.4E 4 Ah Oconee 3 2 motor .13/yr 3/yr .01(f) 6.5E-4 1- 1 turbine ANO 1 1 motor .15/yr 1/yr .061 1.6E-3 IE-3 1 turbine
.14/yr 5.2/yr 1(9I 3.9E-6 7.1E-6 4.4E-2 7E-2 Sion 1 2 motor 2.8E-2 1 turbine IE-6 accapiled using Rets. 2-11.
.bmomenclature defined in Table 2.2.
cMechanical failure in 30 minutes.
dLOSP induced af ter 30 minutes.
eLosp gecovery.
I From Ret. 2.
9No credit taken for MFWS.
hReference'11.
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Table 2.2. Nomenclature Used in Table 2.1
. f(A) Frequency of occurrence of event A
- Q ( A) Probability of failure of system A Q (A/B) Probability of' failure of system A given B has occurred R(A) Non-recovery probability for system A LOSP Loss of.off-site power T1 LOSP transient T2 Loss of MFWS transient M MFWS L AFWS AC Loss of all AC power i
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power will vary significantly depending on the number of trains in the system. For a two-train system which meets the minimum design requirements, the frequency of loss of DC power is on the order of .
10-4 per year. (12) Sequoyah 1, which has four independent trains of DC power, has a frequency of loss of DC power of 1.2 x 10-6 per year.(9) It should be noted, however, that in -
the context of determining the benefits of adding a feed and bleed capability to the CE plants in question, loss of DC power wi'_1 have no impact on the analysis. This is because loss of DC power will not only fail the MFWS and AFWS as noted above, but it will also fail the potential feed and bleed " system". Thus, in the case of loss of DC power, there are no benefits to be gained by having a feed and bleed capability.
A second initiating event which impacts both the MFWS and AFWS but which is not a common mode failure is loss of offsite power.
LOSP will cause loss of the MFWS and put more stringent require-ments on the operation of the AFWS, i.e., it will require operation of the AFWS using the emergency diesel generators or a turbine-driven train. The frequency of occurrence of LOSP is dependent on both electric grid reliability and plant procedures. The frequency of occurrence of LOSP for the plants surveyed ranged from .13 for
, Oconee 3 to .35 for Crystal River 3.
2.3 AFWS Hidden Failures Hidden failures are failures which cannot be detected.by tests, but which can cause system failure. They occur in one of two forms: system misalignment due to test and maintenance and improper system design.
The probability of occurrence of hidden failures due to test and maintenance activities can,be estimated using fault tree techniques. As an example, consider the simplified AFWS shown in Figure 2.1. Valve A must be closed in order to test pump A using test line A. Likewise, valve B must be closed to test pump B or pump C.- If pump A and pump B are tested simultaneously, then the failure to open valves A and 'B af ter the test is a potential com-mon mode failure. -If poor test procedures are used, the failure probability for the AFWS due to hidden failure could be as high as 3 X 10-3 On the other hand, if the pumps are. tested individu-ally by different personnel and good test procedures are used, there is no common mode failure to reopen' valves A and B. Under these conditions, the failure probability for the AFWS due to hidden failure'could be as low as 10-7 or 10-8, The probability of_ occurrence of hidden failures due to l improper system design-cannot be estimated; however, it is assumed ~
- that they do not drive AFWS unreliability. The types of improper design which have been experienced (and.correctad) in the past ,
include improper placement of flow sensors causing the~AFW pumps to be tripped off due to a no-flow signal.(13). Also, trouble
- has been experienced with actuation systems which misinterpret the L opening-of atmospheric dump valves as a main steam line break resultingfin the cut-off of all feedwater. (13)
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2.4 Conclusions on Feedwater Reliability It is necessary in power plants which do not have the -
capability to feed and bleed to have highly reliable feedwater systems in order to respond to transients, SGTR, and some small LOCAs. MFWS reliabilities are relatively consistent from plant to ,
plant, and consequently, it is not necessary to do a detailed analysis of the MFWS. Rather, industry data on MFWS reliability can be used. AFWS reliabilities, on the other hand, vary significantly from plant to plant, and it is necessary to do a detailed analysis for each plant in order to accurately account for system design and potential hidden failures.
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3.0 PORV LOCAs A potentially significant draw-back to installing PORVs on power plants is that they do provide a potential LOCA path. If
' such a LOCA, called a PORV LOCA, occurs, then mitigating systems must respond to prevent core melt. This possibility of core melt could counterbalance the advantages gained by adding a feed and bleed capability to plants. In order to gain a general under-standing of the potential problems created by PORV LOCAs, a survey of PRAs was done. Also, a specific analysis was done to estimate j the probability of a PORV LOCA at San Onofre 2.
3.1 Frequency of PORV LOCAs The estimated frequencies of PORV LOCAs at seven PWRs are presented in Table 3.1. The frequencies vary from essentially zero at Crystal River 3, ANO 1 and Zion 1 to 0.02 per year at l Indian Point 2 and 3. At Crystal River and ANO the PORV set point i is such that all small LOCAs are attributed to the safety valves. I Zion 1 is operated with the PORV block valves closed, thus making all LOCAs induced by a high pressure transient again attributable to the safety valves. At Oconee 3 the frequencies of LOCAs due to PORVs and small breaks of equivalent size are about the same. At Indian Point PORVs are the only sources of small small (S2)
LOCAs considered.
PORV LOCAs can be caused by several initiators.(14) The most predominant initiator is a transient which causes higher than normal. pressures in the reactor coolant system (RCS). This will result in the PORVs being opened by the actuation system. If the PORV fails to close on command and attempts by the operator to use other valves, i.e., block valves, to isolate the PORV fail, a PORV LOCA will ensue. A second and less important PORV LOCA initiator is the spurious actuation of a PORV to the open position with failure to isolate as described above.
3.2 Effects of PORV LOCAs The effects of a PORV LOCA at various PWRs in terms of the release categories used in the Reactor Safety (2) Study are presented in Table 3.2. The release categories are defined in Table 3.3. As can be seen from Table 3.2, the releases-which can result from a PORV LOCA can vary a great deal. The reason for this variance is that the release category depends heavily on the behavior of mitigating systems and containment' systems during the accident.
3.3 Frequency of a PORV LOCA'at' San Onofre 2 The PORV system postulated for use at SONGS 2 would be manually actuated and totally dedicated to providing a feed and bleed capability. It would not be used for mitigating reactor coolant system pressure increases or spikes due to transients.
.g Table 3.1. Data on PORV Initiated LOCAs Contribution of PORV LOCA to Total Plant (a) f(PORV LOCA)I Core Melt Probability f (S 3 ) (b) (c)
Cryst'al' River 3 c c .0013 Calvert Cliffs 2 .018 7.9% .001 ANO 1 e c .0031 Oconee 3' .0015 15% .0013 I
Indian Point 2 .02 4.6% c Indian Point 3 .02 6.7% c Zion l' c c .035 ,
asequoyah'and Surry studies did not address PORV LOCAs.
.b Per year, cFrequency of small LOCA of equivalent size to PORV LOCA.
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1;able 3.2. Probability that'a PORV LOCA whi:h Leads to Core Melt will Result in a Certain Size Rclease i
Release Category Reactor- 1 2 3 4 5 6 7 l
Surry 1% 1% 12% 1% 1% 8% 77%
.: Calve r t - Clit f s ' 2 4% 32% 33% 4% 1% 11% 15% -)
Sequ6yah 1 2%' 20%' 55% 21% 2% -. --
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'Oconee 3 1% 15% 31% 4% 2% 15% 33%
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$- .;ANO 1- 5% 40% 5% 1% 3% 25% 22% ,
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Table 3.3. Release Categories Used in the Reactor Safety StudyI2)
Duration Warning Elevation Containment Time.ot ot Time for of Energy Fraction of Core Inventory Released Release ' Release! Release . Evacuation Belease Release Category (Hr) (Hg) (Hr) .(Meters) (106 Btu /Hr) Xe-Nr Org. I 1 Cs-Rb Te-Sb Ba-Sr Bu(a) La(b) l' 2.5 0.5 1.0 25 520 ICI 0.9 6E-3 0.7 0.4 0.4 0.05 0.4- 3E-3 2 2.5 0.5 1.0 0 170 0.9 7E-3 0.7 0.5 0.3 0.06 0.02 4E-3 3 5.0 1.5 2.0 0 6 0.8 6E-3 0.2 0.2 0.3 0.02 0.03 3E-3 g 4 2.0 3.0 2.0 0 3 . 0. 6 2E-3 0.09 0.04 0.03 5E-3 3E-3 4E-4
- Fd 5 -2.0 4.0 1.0 0 0.3 0.3 2E-3 0.03. 9E-3 5E-3 IE-3 6E-4 7E-5 PJ g 6 12.0 10.0 1.0 0 N/A 0.3 2E-3 8E-4 8E-4 IE-3 9E-5 7E-5 IE-5 7 10.0 10.0 1.0 0 N/A 6E-3 2E-5 2E-5 IE-5 2E-5 lE-6 1E-6 2E-7 alncludes No, Rh, Tc, Co.
b l ncludes No, Y. Ce, Pr, La, Nb, Aa, Co, Pu, Np, 3r.
-cA lower energy release rate than this value applies to part of the period over which the radioactivity is being released.
The ef fect et lower energy release rates on consequences is found in Appendix VI of Ref erence 2.
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Also, the set point of the PORVs would be the same as for the safety valves. Therefore, assuming appropriate management controls were used to prevent accidental PORV opening by an e operator, e.g., the PORV Gwitches were locked in the closed position, the possibility of a PORV LOCA would be attributable to
. incorrectly setting the PORV set point. The probability of this type of error is estimated at 3 x 10-3,(15) Taking credit for block valve operation and accounting for two PORVs leads to an estimated frequency of a PORV LOCA at SONGS 2 of 1.7 x 10-7 per year (see Figure 3.1). This LOCA frequency is smaller than most of those in Table 3.1, because an additional operator error, failure to correctly set PORV set point, is required. Since the frequency of PORV LOCAs at SONGS is so low, they can be considered to have a negligible impact on plant risk and will not be considered in the analysis in Section 4.
3.4 Conclusions on PORV Related LOCAs A significant draw-back to adding PORVs to a power plant is that they provide a potential LOCA path. If the PORVs are to be used sclely for feed and bleed, as would be the case at SONGS 2, as proposed here, the frequency of such a LOCA is about 1.7 x 10-7 per year. The potential radiological consequences of a PORV LOCA are the same as for any small LOCA. They can be relatively minor or extremely severe, depending on the success of the safety systems required to respond to the LOCA.
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4 LOCA LOCA THROUGH THROUGH PATH 1 PATH 2 I
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- FAILURE PROBABILITIES AND FREQUENCIES ARE LISTED IN TABLE 3.4 Figure 3.1: Fault Tree for Determining the Frequency of a POBV LOCA at San Onofre 2 r
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I OPERATOR BLOCK VALVE I O V4 FAl' S TO ISOLATE LOCA CF z 2 Figure 3.1: (Continued)
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Table 3.4. Data Used to Determine PORV LOCA Frequency I
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I Event Freauency Probability Source j - High Pressure .278/yr 14 Transient l'
{ PORV Set Point 3.OE-3 15 i Incorrect I PORV Fails to .02 18 l -Reclose 1
! Block Valve Fails .001 18 l to Close Operator Fails to 3.OE-3 15 j . Isolate LOCA l
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4.0 BENEFITS OBTAINED BY ADDING A FEED AND BLEED CAPABILITY TO SAN ONOFRE 2 The approach taken in this study to estimate the benefits to be obtained by adding a feed and bleed capability to the San -
Onofre 2 power plant was to determine the change in core melt frequency brought about by such a capability. To perform the analysis, alterations were made in fault tree models of the Calvert Cliffs 2 power plant which were developed in TAP A-45 to reflect system differences between the San Onofre and Calvert Cliffs plants. These alterations included developing new AFWS, electrical system, and Auxiliary Spray System models for SONGS and altering the suction lines and valve alignment of the HPSI system of the Calvert Cliffs model. The net result of these changes is a model which contains the front line systems and electrical system of the San Onofre plant and the support systems, e.g., component cooling water system, of the Calvert Cliffs plant. The Calvert Cliffs support systems models allow common mode failures to be modeled. (It should be noted that the major contributors to failures in the support systems are pump fdilures. Consequently, differences in the piping arrangements of the two plants' support systems have a minor impact on the results of the analysis.)
In order to present the information used in and obtained from this analysis in an orderly and concise manner, the following for-mat is used. First, the results of the analysis of the SONGS AFWS are presented and compared to an earlier, similar analysis.(10)
Next, the event trees used in the analysis, which reflect the capabilities of SONGS to respond to certain internal events both with and without a feed and bleed capability are discussed. And third, the results of the analysis are presented.
4.1 Reliability of the San Onofre 2 AFWS The reliability estimated for the SONGS AFWS, Figure 4.1, under various conditions is given in Table 4.1. This table also compares the reliability found in this study to the reliability reported in a study done by CE.(10) The reliabilities found in this study and the CE study compare relatively well except for the case of loss of offsite power (LOSP). This study estimated the AFWS unreliability given LOSP to be a factor of about'10 higher than the CE study. The apparent reason for this difference is in-l the modeling. Specifically, the CE model does not appear to i
account for failure of an AFWS actuation train due to loss of its supporting DC power train. As a result, it seems that the CE model does not then account.for failure of the turbine-driven AFWS ,
train due to loss of either DC power train.
4.2 Event Trees .
Three cases of plant capabilities are considered here to
' establish the benefits to be'obtained by adding a feed and bleed capability to the San Onofre 2 power plant. The first case is the present design which does not include a feed and bleed capability.
FL w,
4730 A p X TO SG 088 CST J T-121 P504 4705)M V"FC T
- pt 04
_ 4714 FO p140 ~
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8 FROM 4715 CG 089 ' 4706)M TO FL j
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, , FC 4731 Figure '4.1: Simolified Drawing of the SONGS 2 AFMS
_ - - -. ~
Table 4.1. Reliability of the San Onofre 2 AFWS i AFWS AFWS AFWS
- Reliability Reliability Reliability Initiating Without with Recovery CE Safety (10) .
Event Recovery (50 min.) (Without Recovery)
Loss of Offsite 5.4E-4 8.0E-5 3.8E-5 Power Loss of MFWS 5.6E-5 2.2E-6 2.2E-5 i
Loss of All AC 2.3E-2 4.9E-3 2.0E-2 4
Power i
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The second case is the addition of PORVs which allows for fecd and bleed using the present HPSI system, Figure 4.2. The third case
, is the addition of a single train system which would be used solely for feed and bleed, Figure 4.3. These last two cases are discussed in detail in Appendix A.
The transient and SGTR-small LOCA event trees for San Onofre as it is presently designed are shown in Figures 4.4 and 4.5, respectively. In the event of a transient, the MFWS and AFWS are the two plant systems considered for removing decay heat from the reactor coolant system (RCS). No credit is given for depressuriz-ing the steam generators and using a condensate pump to supply feedwater, although it is a possible option. If both the MFWS and AFWS fail, core melt will ensue (Sequence 3). In the case of a SGTR or small LOCA, either the MFWS or AFWS must remove decay heat from the RCS, and inventory make-up is provided by the HPSI sys-tem. Core melt will occur if any one of the Sequences 5, 7, or 8 occurs.
The transient and SGTR-small LOCA event trees which reflect the addition of PORVs to San Onofre are shown in Figures 4.6 and 4.7, respectively. For transients, decay heat can be removed from the RCS using either the MFWS or AFWS. If both of these systems fail and neither can be recovered within 20 minutes, the core can be cooled by going to a feed and a bleed mod'e of operation. (If feed and bleed fails, there is still an additional 30 minutes for recovery of the MFWS or AFWS based upon the CE thermal-hydraulic analyses.) In the case of a SGTR or small LOCA, there are several options available to prevent core melt. Heat can be removed from the RCS via the MFWS or AFWS with make-up being provided by the HPSI system, or if the MFWS and AFWS both fail, feed and bleed can be used to cool the core.
The transient and SGTR-small LOCA event trees for the case of a dedicated feed and bleed train added to San Onofre are shown in Figures 4.8 and 4.9. Given a transient, three systems, the MFWS, AFWS and feed and bleed system, are available to prevent core melt. (The feed and bleed system would be initiated 40 minutes into the accident if necessary based upon the B&R analyses reported in Appendix A.) In case of a SGTR or small LOCA, the plant has response capabilities similar to the above case. However, with the dedicated system there are 40 minutes rather than 20 minutes available for recovery of the MFWS or AFWS before the feed and bleed system would have to be initiated.
It should be noted that SGTR scenarios which involve the
, opening and failure to close of a steam generator safety valve are not considered in this analysis. Such scenarios result in a LOCA outside of containment. (Water from the reactor coolant system flows through the tube rupture to the secondary system and escapes to the environment through the safety valve.) However, the LOCA can be terminated by operator action before the refueling water tank (RWT) is depleted about 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> into the accident. Once the LOCA has been terminated, decay heat can continue to be removed by the unaffected steam generator.
l
-- ____ ______________-_-____a
M 1-C-724 CWT i PUMP MV-932 3 My-9301 12-C-358 M 6-C-725 18-A-551 7-C-212 y RAS toop 1 A
- MV-9324 27-A-551 M 3-C-724 "
,s ,T MV-9303 My-9326 M 19-A-551 pq LOOP 1B L.c.;;i3-C-o75 Mv-9 27 29-A-551 L.C.}(10-C-212 3 16-C-358 I h ,s I
y d L.O. M V-93 29 F
L*C*n'11-C-212 L.C.][14-C-075 M 20-A-551 I LOOP 2 A MV-9330 31- A-551 M 2-C-724 es RWT PUMP MV-9332 MV-9300 8-C-675 15-C-358 C-553 21-g-C-212 y
ras N N U t.O.
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- simplified Drawing Of the SONGS 2 Epg1 system
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IE* RPS MFW AFW Result Sequence Fi S 1 K
E S 2 M
i " y, N, CM 3 i
K ___
f The nomenclature used in the event trees along with scme assumptions and data are found in Table 4.2.
Figure 4.4: -Transient Event Tree for San Onofre 2 Without PORVs 4 8 0 e
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Table 4.2. Nomenclature. Assumptions, and Data Used in the San Onofre Event Trees Definition Notes Event IE initiating event TE transient event RPS.K reactor protec- RPS failure was not considered in tion system this study.
MFW,M main feedwater *Q(M/T3 ) = SE-3 (from Calvert system Cliffs data)
Q(M/SGTR or small LOCA) = 0.l(15)
If AFWS failure occurs, it is
' assumed all resources are use in AFWS recovery, and so MFWS recovery is not considered.
AFW.L auxiliary feed-water system HPSI.D high pressure safety injection system GR/VO.P relief valve Q(P) = SE-3(15.18) opens F+B.F feed and bleed Includes PORV failure system T1 loss of off- f(T I) = 0.09(11) site power T2 loss of MFWS f(T 2) = 1.2 T3 transient not f(T 3) = 7.5(16) causing initial loss of NFWS SGTR f(SGTR) = 0.02, based on a conserva-
.tive calculation using operation data from CE plants Small f(S2) = 0.02(17)
LOCA S2
- Q() = probability of failure.
- f() = frequency of occurrence (per. year).
IE RPS MFW AFW HPSI Result Sequence E
S 4 Ti D CM 5 K D S 6 1
T '
D CM 7 M
Anmp S LOCA 2
L CM 8 t
E _-
Figure 4 5 : SGTR - Small LOCA Event Tree for San Onofre 2 Without PORVs
IE RPS MFW APW SR/VO HPIS Result Sequence Ei S 9 K L s lo D 11 M S P
t e
P CM 13 E
Figure 4.6 : Transient Event Tree for San Onofre 2 With PORVs
IE RPS MFW AFW SR/VO IIPSI Result Sequence
_D g 14 Ei U CM 15 h D S 16 K
L D
CM 17 t
kV cm i
D S 18 P
SGTR S LOCA-
- D 2 CM 19 L t l
t P
CM 20 K
L Figure 4.7: SGTR - Small LOCA Event Tree for San Onofre 2 With PORVs
IE BPS MFW AFW F&B Resu1t Sequence M
S 21 l
ii -
L S 22 M
i M P y S 23 TE L
F CM 24 K _,_
Figure 4.8: Transient Event Tree for San Onofre 2 With a Dedicated Feed and Bleed Systcm
IE RPS MFW AFW HPST FEB Result Sequence E 25 s
~R F 3 26 D
F CM 27 E b 28 s
E I E w s 29 o
.l D F --
CM 30 ccen S2 'UCA -
F s 31 L
F 32 CM K' _-
Figure 4.9 : SGTR - Small LOCA Event Tree for San Onofre 2 With a Dedicated Feed and Bleed System G e & G
l 4.3 Analysis Results The results of the analysis of the three cases of plant cap-abilities for San Onofre are presented in Tables 4.3-4.6. In summary, the core melt frequency at San Onofre due to transients, SGTR, and small LOCAs could be reduced by a factor of 1.3 (4.0E-6 per year) by adding PORVs and a factor of 3 (1.2E-5 per year) by adding a dedicated, single train feed and bleed system.
For the case in which only PORVs are added and feed and bleed is accomplished using the HPIS, there is a very small relative 1
change in the core melt frequency. The reason for this is that no significant benefit is obtained for SGTR or small LOCAs. The SGTR and small LOCA accident sequences for Case 1 which involve failure
. of the HPIS (Sequences 5 and 7) cannot be recovered using feed and bleed, because the HPSI system has already failed. Also, the other core melt sequence which does not involve HPSI failure (Sequence 8) makes only a small contribution to the overall core melt frequency, and thus recovery of this sequence using feed and bleed has a very small impact.
For the case in which a dedicated, single train feed and bleed system is added, the relative reduction in the core melt frequency is better than in the previous case; a factor of 3 compared to a factor of 1.3. As can be seen from Table 4.6, the reduction in core melt for transient sequences obtained with the proposed feed and bleed system is not as large as for the case when PORVs only are added. The reason for this is that the feed side of the feed I and bleed system is not redundant, whereas the HPSI system does
! have redundancy. For the SGTR-small LOCA sequences, however, there is a significant reduction in the core melt sequence frequencies (a factor of 100). This is because the plant has the capability to feed water to the RCS even though the HPSI system may have failed.
4 4
4 0
Table 4.3. Core Melt Frequencies for. San Onofre 2 Without PORVs (Case 1)
Sequence Core Melt Number Sequence Frequency (per year) ,
3 T1ML 7.0E-6 3 T2ML 2.6E-6 3 T 3 ML 8.3E-8 Subtotal 9.7E-6 5 S2RRD 3.8E-6 7 S2RMED 3.8E-7 8 S2KML 4.4E-9 Subtotal 4.2E-6 5 (SGTR) RED 3.8E-6 7 (SGTR)RMED 3.8E-7 8 (SGTR)KML 4.4E-9 Subtotal 4.2E-6 Total 1.8E-5 e
t l.
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Table 4.4. Core Melt Frequencies for San Onofre 2 With PORVs (Case 2)
Sequence Core Melt Number Secuence Precuency (Der year) 12 T 1RMLED 5.5E-6 12 T 2RMLPD 7.2E-8 12 T 3RMLPD 2.3E-9 13 T tRMLP 3.6E-8 13 T 2RMLP 1.3E-8 13 T 3RMLP 4.2E-lO Subtotal 5.6E-6 15 S2RRD 3.8E-6 17 S2RMED 3.8E-7 19 S2RMLPD 1.2E-lO 20 S2RMLP 2.2E-ll Subtotal 4.2E-6 15 (SGTR) RAD 3.8E-6 17 (SGTR)RMED 3.8E-7 19 (SGTR)RMLPD 1.2E-lO 20 (SGTR)RMLP 2.2E-ll Subtotal 4.2E-6 Total 1.4E-5 O
Table 4.5. Core Melt Frequencies for San Onofre 2 with a Dedicated Feed and Bleed System (Case 3)
Sequence Core Melt Number Sequence Frequency (per year) ,
24 T IRMLF 5.9E-6 24 T 2RMLF 6.7E-8 ,
l 24 T 3RMLF 2.lE-9 Subtotal 5.9E-6 26 S2RRDF 3.8E-8 29 S2RMEDF 4.OE-9 32 S2RMLF 1.lE-lO Subtotal 4.2E-8 26 (SGTR)RNDP 3.8E-8 29 (SGTR)RNUDF 4.OE-9 32 (SGTR)RMLF 1.lE-lO Subtotal 4.2E-8 Total 5.6E-6
i Table 4.6. Comparison of Core Melt Frequencies at San Onofre with and Without A Feed and Bleed Capabilit; Frequency -Frequency Frequency Sequences Without PORVs With PORVs With F+B System Transient 9.7E-6/yr 5.6E-6/yr 5.9E-6/yr Small LOCA 4.2E-6/yr 4.2E-6/yr 4.2E-8/yr SGTR 4.2E-6/vr 4.2E-6/vr 4.2E-8/yr Total 1.8E-5/yr 1.4E-5/yr 5.6E-6/yr i
5.0 COSTS ASSOCIATED WITH INSTALLING A FEED AND BLEED CAPABILITY This section discusses the feasibility and costs of implement- .
ing controlled depressurization of the primary system in PWRs using PORVs. Such a capability may be required for certain acci-dent conditions, such as total loss of feedwater flow (TLOFW),
where there is no major loss of primary coolant that would result in rapid depressurization to the level at which the high-pressure safety injection flow is inittated.
5.1 Study Methods Two cases were investigated. The basic. case considers adding two large PORVs, with relieving capacity well in excess of that required for decay heat removal, to depressurize the system as rapidly as possible to permit an existing HPSI pump to initiate flow injection. Valve capacity and time of opening after TLOFW for this case were based on the results of earlier transient analyses performed by CE.
The second case investigated involved of upgrading both the feed and bleed capabilities by supplementing the existing HPSI pumps to permit flow injection to be initiated at or near full system pressure, in addition to adding PORVs. Investigation of this more conservative case was based on two factors:
- 1) Uncertainties in the final definition of the criteria for protection against core uncovering; i.e., whether the requirement is that the top of the two-phase mixture zone remain above the top of the core, or the more conservative requirement that collapsed water level must remain above the top of the core. The latter would require an upgraded HPSI capability.
- 2) The uncertainties about the modeling used for transient analyses and interpretation of the results obtained.
In considering upgrading of the HPSI pumping capability the flow rate considered was that necessary to remove decay heat by water heatup only; and at the latest time after TLOFW for which the CE transient analysis showed that feedwater could be restored without core uncovering. Approximate calculations, which included allowance for the transit time for the cool injected water to reach the core, indicated that this injection flow rate would lead to core uncovering as defined by the collapseo water level. There-fore, two additional high injection flow rate cases at or near .
full primary system pressure were considered which would prevent the collapsed water level from dropping below the top of the core:
I
- 1) Early initiation of HPSI flow at a few minutes after TLOFW.
2)' Initiation of_HPSI flow at the latest possible time after TLOFW for which auxiliary feedwater flow could be restored.
The latter of the above two cases was used to develop a conceptual design based _on the most stringent requirements for pump _ flow and electric power. It is expected that ultimate selection of pump size based on more definitive core uncovering criteria and more detailed transient analyses would result in a flow requirement-lying below the bounding value used.
Transient analyses made by CE in early 1983 have a relationship to this study in the following respects:
o The total relieving capacity of the PORVs here is the same as that used by CE.
o Water level information from the TLOFW transient analysis was used as an input condition to determine upgraded HPSI flow requirements to prevent core uncovering on a collapsed water level basis. r o -With new large PORVs and an existing HPSI pump the vessel two-phase mixture level drops to about one-third below the top of the core, which tends to substantiate consideration of an upgraded HPSI pumping capability as an alternative to depending on an existing pump.
o The CE transient analysis'was based on-no loss of offsite power with -the reactor being tripped amt 30-seconds and the. reactor ~ coolant pumps at 10 minutes after TLOFW._ This study also took into' account the possibility of a. loss of offsite power during a TLOFW condition and the requirement-that the feed and bleed l system would tave to be powered from an emergency
! diesel generator.-
l
! Results of transient analyses performed by Argonne National!
4 Laboratories (ANL) for conditions similar to those of the CE analysis indicate that the two-phase-mixture' level may not drop.
below the top of the core. These_results indicate 1that TLOFW with i no: loss of offsite power-may not require 1an upgraded HPSI. pumping.
. capability and that TLOFW combined with-loss'of offsite power is
- less.likely to require this, i
For each Lof 'the two -cases considered two subcases- w'ere:
I. + --inves tig a ted :
L ~.
--37 -
- ._ _ _ c. _ _ _ . -_ . -
- 1) Installation in a new plant during the final stages of its construction.
- 2) Installation in a plant that had been operating for -
some time.
The San Onofre Nuclear Generating Station was selected as a plant not currently having a feed and bleed capability to deter-mine the feasibility and costs of implementing such a capability.
At the same time the implemention of the upgrade in other plants of similar design was examined to determine what aspects of the design could make a significant difference on a plant-specific basis.
5.2 Conceptual Design for Plant Upgrading A conceptual design was established for a complete feed and bleed system embodying the following features:
o Capability of being retrofitted in SONGS.
o A bleed capability provided by large PORVs consistent with the CE transient analysis, o A feed capability which would assure no core uncovering, based on the collapsed water IcVel, at a sufficiently long time after TLOFW to ascertain if auxiliary feedwater flow could be restored.
The single conceptual design actually covers the two basic cases of new large PORVs used in combination with an existing HPSI pump, and the complete feed and bleed case in which the HPSI pumping capability would be upgraded.
A full set of design criteria'has been developed for the system design. The more important of these criteria include the requirement that the system equipment and piping must be 1
consistent with the existing safety inje~ction systems with respect to ASME - Code Class, Nuclear Safety Class, Quality Group Class and Seismic Category; the new PORVs must be fully safety grade and environmentally qualified; the system must be capable of operation when offsite and turbine-generator power sources are unavailable (e .g . , from a single existing diesel-generator); the new system must in no way affect the functions of the existing safety injection systems; and redundancy is not required (e .g . , the single-failure criterion does not apply).
Bleed side mechanical design consists of the two PORVs mounted '
at the pressurizer top using the nozzles provided for the existing safety relief valves (SRVs), c quench tank (similar to the
[ existing quench tank), and connecting piping.
i l Feed side mechanical design includes the new HPSI pump with its suction taken from the refueling water storage tanks (RWSTs) l
~
and its discharge connected to the four low-pressure safety injection (LPSI) lines to the reactor coolant loops at a point
. where the existing lines have changed rating to the high pressure level of the reactor coolant system. This connection is outside of primary containment so no new piping penetrations are required.
Consideration was given to possible upgrading of an existing HPSI pump or the addition of a booster pump. Both of these were found not to be feasible; therefore, a completely separate pump capable of injecting at full reactor coolant system pressure has been specified. Room was found for this pump in the SONGS compartment containing the No. 2 HPSI and LPSI pumps; however, neither of the other two pump compartments would have room for a redundant pump if the single-failure criterion were imposed.
The main consideration in the electrical design is the ability to supply power to the new, large HPSI pump from an existing diesel-generator. If neither the existing HPSI or LPSI pumps would be required or allowed to run while the new HPSI pump was in operation, operation would be possible with one SONGS diesel-generator but with only a small reserve margin with respect to the machine's nameplate rating. Supply of electrical power for the first case where an existing HPSI pump is used poses no problem.
No major structural changes or additions would be required to accommodate the feed and bleed system. Structural work would consist mainly of additional pipe supports, pump base, platforms, walkway. And railings.
In the present conceptual design the basic control for Cases 1 and 2 is automatic; that is, the PORVs open and the new HPSI pump starts in response to a higher than normal primary system pressure and a low feedwater flow signal. If manual actuation was used, there would be a corresponding decrease in installation costs.
In Case 1 the PORVs would be fully opened to drop the pressure as rapidly as possible to the level where an existing HPSI pump would initiate flow. In the full feed and bleed, Case 2, the PORVs open and the new HPSI pump starts in response to the same high pressure and low feedwater flow signals. However, the PORVs would function in a modulating mode to hold pressure constant until decay heat generation rate is low enough (about 3-1/2 hours) to permit the shutdown cooling mode (SDC) of the LPSI pumps to take over. At this time the PORVs would be fully opened and the new HPSI pump stopped.
In the final design of a control system, consideration may be
. given to an all-manual control system for the following reasons:
o Simplicity of control and avoidance of spurious initiation.
o Elimination of the need to interface with existing primary pressure and feedwater flow instrumentation channels; thus, no possibility of jeopardizing these channels.
o Lower implementation time and costs.
o Operator choice to maintain full primary pressure to the latest possible time after TLOFW for possible -
restoration of auxiliary feedwater for cooling.
5.3 Detailed Design, Installation and Operation of a Feed and Bleed Capability The detailed engineering and design, based on the conceptual design developed in this report, would be of the type normally performed for nuclear power plant safety systems. Because of the expectation that a system for a particular plant would either be designed and installed during the latter stages of overall plant construction, or retrofitted to an operating plant, the engineering and design would have to be organized as a separate project with a dedicated project team. This would help keep the system installation off of the critical path.
The nature of the system application will require a significant amount of special analysis, such as:
o Additional thermal-hydraulic transient analyses to establish whether the system can meet criteria.
o Electrical load analyses to define required loads and evaluate ability of existing system to supply them.
o Analyses of actual radiation levels for controlling personnel exposure, o Stress analysis due to added loads on critical' piping, o Analytical studies to support actual valve selection.
o Analyses to consider the effects of solid water operation, cold repressurization and primary system cool down rates.
I Coordinated schedules for engineering, design and analysis and construction have been developed. The former has a span of 18 months and the latter 12 months. There is a six month overlap resulting in an overall project schedule of 24 months. The schedule is keyed to an annual outage for refueling and scheduled maintenance, which is assumed to be of two months duration. Start of project work would be dependent on NRC requirements having been l established. The major critical path item of the schedule would
( be the new HPSI pump procurement, which is expected to require .
l 13-14 months. Installation of the system within the planned time l would depend to a large extent on. careful planning.
In retrofitting a feed and bleed system to a plant that has been in operation for some time, radiation exposure to personnel will be a concern. The two potential problem areas involved are around the pressurizer within containment and adjacent to the LPSI pump in the pump cubicle where data from several plants that have
, been operating for a number of years indicate that shutdown radia-tion levels can be as high as 0.3-0.4 R/hr. Such levels would severely limit the time that personnel could spend in the area during installation. In the case of the pump area, the use of temporary shielding should attenuate radiation levels sufficiently to eliminate any need to limit occupancy time. Since the use of temporary shielding in the pressurizer area may not suffice, it has been conservatively assumed that allowance would be required for burnout of installation personnel.
Although the feasibility of installing a system for providing feed and bleed capability was investigated specifically for SONGS, the conceptual design and other material developed in this report would have a generic applicability to other plants lacking this capability. Factors that could be expected to affect the feasibility for a specific plant would include:
o The arrangement of equipment and piping around the pressurizer and the availability of a suitable connection for the PORVs.
o Room for the installation of a new HPSI pump if required.
o Accessibility of safety injection piping for the necessary interconnections.
o Capacity of the existing emergency diesel generators to carry the additional electrical load.
Once a PORV type has been selected and qualified for the l application in the first plant, it should be suitable for use in any subsequent plants being upgraded. This is also true of a new HPSI pump, if found to be required.
5.4 Implementation Costs Estimates to cover the complete cost of implementation for the two basic cases, addition of PORVs with an existing HPSI pump and addition of PORVs with a new HPSI pump, were made for installation in a new plant under construction and a plant that had been in operation for some time. The total estimates for each of these cases are shown below:
o Case lA - PORVs only, new plant $2,495,000 ,
o Case 1B - PORVs only, operating plant $4,254,000 o Case 2A - PORVs and pump, new plant $6,958,000 o Case 2B - PORVs and pump, operating plant $10,310,000 Construction costs and costs for supporting services were
, estimated separately. Prevailing construction labor rates in the San Diego area were used and allowances made for three shift -
operation, premium time on weekends, overtime at shift changes for work during the schedule plant outage, and travel allowance for construction workers. In the case of installation in an operating plant, allowance was made for the additional manhours and other costs associated with burnout of craft labor personnel in high radiation areas and also for the general difficulties associated with working in an operating plant.
Present day costs were used and escalation applied at six percent / year using the developed schedule. Allowance was made for interest during construction at an annual rate of 12%. An overall contingency allowance of 25% was made. The costs are detailed in Appendix A, which is the Burns & Roe report.
There are two cost factors not included in the present estimates which could have a very significant effect in another plant-specific situation:
- 1. If the addition of new HPSI pump is required, the additional electrical load may not permit use of an existing emergency diesel generator. Costs associated with the addition of a dedicated diesel-generator for the feed and bleed system, including structures and supporting systems, could be approximately $11,000,000 which is somewhat greater than the total ~of all other implementation costs. This would not be a concern where PORVs only would be added.
- 2. In the case of an operating plant, replacement energy costs incurred by prolonging a_ scheduled annual outage by the installation of the system could result in costs that would exceed the total of all other implementation costs in just a few days, considering that replacement energy costs are typically in the range of $500,000 to $1,000,000 per day. In an actual installation, if the work could not be all completed in the period of one annual outage, it could be completed during the following year's outage if NRC l rules permitted. The necessity for hydrotesting at the completion of system installation may extend the outage by two or three days; however, as is often the
- case, the turbine-generator work may be on the critical path in determining the outage time.
Controls-for the full feed and bleed system were estimated on l- the basis of automatic control. The alternative use of manual l controls would reduce the costs by approximately $230,000; '
i however, costs would not be a primary' consideration in selecting cutomatic versus manual control.
4 i 5.5 Conclusions on Feasibility and Costs
. The following general conclusions have been reached as the result of the cost study performed:
. o For PWR plants lacking the capability, addition of ,
PORVs to permit controlled depressurization would be
- feasible.
o Transient analyses performed by CE tentatively indicate that the use of PORVs with an existing HPSI pump may prevent core uncovering. However, a final decision on
- whether an upgraded pumping capability will be required can not be made until firm criteria for fuel protection l
, have been established in terms of amount and type (two-
{. phase mixture vs. collapsed water level) of core uncovering allowable and/or maximum fuel clad i temperature. Further transient analyses may be l necessary to reach this decision and to establish the i required injection flow rate if upgrading of pump
! capability is found to be necessary.
o For purposes of developing a conceptual design and investigating the feasibility of implementation, an i
upper bound case of injection flow rate and associated pump size has been determined on the conservative basis of no core uncovering in terms of collapsed water level 4
and the requirement to provide injection at full system pressure. Depending on the results of further-transient analysis, it may be possible to meet the 4
conservative criterion for-no core uncovering by use of i'
a new pump having head and flow requirements less than those of the' pump specified in this_ study but in excess of the existing HPSI pump capability.
o A conceptual design that is applicable to either the
- full feed and bleed system or to the use of PORVs only has been developed. This could serve as the starting point for proceeding with detailed engineering and design for a specific plant.
, o Detailed engineering and design for a specific. plant would have to be supported by.a significant amount of specialized analysis.
~
o . Installation of a system would have to be very carefully planned and_ executed,-particularly in an operating plant.. An overall schedule of two years is considered feasible. Keying of schedule to-an annual
. scheduled, outage would be. essential.
I o--Radiation exposure to personnel for installation in an.
operating plant will-have to be taken into account but
.' appropriate allowances;can be made.
> 43-
o The conceptual design developed is generally applicable to other PWR plants requiring the depressurization capability. However, certain factors such a diesel-generator capacity, equipment layout.and piping arrangement will strongly influence the feasibility and costs for a given plant. ,
o Implementation costs range from $2.5 million for PORVs only in.a new plant to $10 million for a complete feed and bleed system in an operating plant. Inclusion of a dedicated diesel-generator or the incurring of replace-ment energy costs for extended outages could increase these estimates very significantly.
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6.0 COST / BENEFIT ANALYSIS Table 6.1 contains an overview of the benefit and cost infor-mation from Sections 4 and 5, respectively. If PORVs were added
. to San Onofre 2 before operation, the core melt frequency would be reduced by a factor of 1.3 (an incremental reduction of 4.0E-6) for a cost of $2.5M. If they were added after operation during a routine outage, the cost would be $4.3M. If a dedicated feed and bleed system were added before operation, the frequency of core melt would be reduced by a factor of 3 (an incremental reduction of 1.2E-5) for a cost of $7.0M. If it were added after operation during a routine outage, the cost would be $10.3M.
O
1 Table 6.1. Cost / Benefit Comparison of Adding a Feed and Bleed Capability to the San Onofre 2 Nuclear Power Plant Feed and Bleed l PORVs Only System Frequency of Core Melt 1.4E-5/yr 5.9E-6/yr Reduction in Core Melt 4.0E-6/yr 1.6E-5/yr Frequency (factor of 1.3) (factor of 3)
Cost (New) $2,495,000 $6,958,000 Cost (After Operation) $4,254,000 $10,310,000 I
a I [
REFERENCES
- 1. Letter from R. L. Tedesco (NRC) to A. E. Scherer (Combustion Engineering), "Depressurization and Decay Heat Removed
- Capability of the CESSAR Design," Docket No. STN 50-470, March 26, 1982.
- 2. U.S. Nuclear Regulatory Commission, " Reactor Safety Study - -
An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," WAfH-1400 (NUREG-75/014), October 1975.
- 3. A. A. Garcia, et al., " Crystal River-3 Safety Study,"
NUREG/CR-2515, SAND 81-7229/I, December 1981.
- 4. S. W. Hatch, et al., "RSSMAP: Calvert Cliffs #2 PWR Power Plant," NUREG/CR-1659/3 of 4, SAND 80-1897/3 of 4, May 1982.
- 5. G. J. Kolb, et al., "RSSMAP: Oconee #3 PWR Power Plant,"
NUREG/CR-1659/2 of 4, SAND 80-1897/2 of 4, May 1981.
- 6. G.J. Kolb, et al., "IREP: Analysis of the Arkansas Nuclear One-Unit 1 Nuclear Power Plant," NUREG/CR-2787, SAND 82-0978, June 1982.
- 7. Consolidated Edison, Power Authority of the State of New York," Indian Point Probabilistic Safety Study," Spring 1982.
- 8. Commonwealth Edison, " Zion Probabilistic Safety Study," Fall 1981.
- 9. D. D. Carlson, et al., "RSSMAP: Sequoyah #1 PWR Power Plant," NUREG/CR-1659/1 of 4, SAND 80-1897/1 of 4, February 1981.
- 10. C. A. Whitaker, Jr., " SONGS-Units 2 and 3: Simplified AFW Reliability - Fault Tree Analysis," Calculation Number S-PEC-246, November 1980.
- 11. A. S. McClymont, et al., " Loss of Offsite Power at Nuclear Power Plants: Data and Analysis," EPRI-NP-2301, March 1982.
- 12. P. W. Baranowsky, et al., "A Probabilistic Safety Analysis of DC Power Supply Requirements for Nuclear Power Plants,"
NUREG-0666, April 1981.
- 13. J. W. Minarick, et al., " Precursors to Potential Severe Core Damage Accidents: 1969-1979 A Status Report," NUREG/CR-2497, June 1982.
- 14. Combustion Engineering, "PORV Failure Reduction Methods -
Final Report," CEN-145, December 1980.
- 15. A. D. Swain, et al., " Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications,"
NUREG/CR-1278, April 1980. ,
- 16. A. S. McClymont, et al., "ATWS: .A Reappraisal - Part 3:
Frequency of Anticipated Transients," Electric Power Research .
Institute, EPRI NP-2230, January 1982.
- 17. Memorandum for D. G. Eisenhut, NRC, from T. E. Murley, NRC,
Subject:
Reactor Coolant Pump Seal Failure, nd.
- 18. D. D. Carlson, et al., " Interim Reliability Evaluation i Program Procedures Guide," NUREG/CR-2728, SAND 82-1100, January 1983.
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APPENDIX A 4
l A STUDY OF CONTROLLED DEPRESSURIZATION
_ OF THE PRIMARY-SYSTEM IN PRESSURIZED WATER REACTORS I
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TABLE OF CONTENTS Page
1.0 INTRODUCTION
............................................ A-8 .
2.0 EXECUTIVE
SUMMARY
....... ................................ A-13
- 3.0 FEED & BLEED METHODS FOR CONTROLLED DEPRESSURIZATION .... A-24 3.1 Methods Considered ................................. A-24 3.2 Method Selected for Conceptual Design .............. A-30 4.0 CONCEPTUAL DESIGN ............. .......................... A-40 4.1 Design Criteria .................................... A-40 4.2 System ............................................. A-42 4.3 Components ....................................s.... A-47 4.4 Arrangement of Equipment and Piping ................ A-54 4.5 Electrical ......................................... A-58 4.6 Instrumentation & Control .......................... A-64 4.7 Structural ......................................... A-71 5.0 IMPLEMENTATION .......................................... A-75 5.1 Installation ....................................... A-75 5.2 Testing ............................................ A-76 5.3 Costs .............................................. A-78 5.4 Schedule ........................................... A-84
- 5.5 Radiation Exposure ................................. A-89 l
OPERATION AND MAINTENANCE ............................... A-91 6.0
(
l 6.1 Operation .......................................... A-91 6.2 Maintenance ........................................ A-91 ~
6.3 Testing ............................................ A-92 .
- 7.0 REQUIREMENTS FOR DETAILED ENGINEERING, DESIGN AND ANALYSIS ................................................ A-93 A-2
I TABLE OF CONTENTS (Cont'd)
Page a
j
. 8.0 APPLICABILITY TO OTHER PLANTS ........................... A-96 ,
- REFERENCES ................................................... A-98 i
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'A-3, i.
LIST OF FIGURES Page .
- 1. Flow Diagram - Feed & Bleed System ...................... A-43
- 2. Pump Outline ..................~..........................,A-43
- 3. Bleed Valve Outline ..................................... A-51
- 4. General Arrangement - Feed Side ......................... A-55
- 5. General Arrangement - Bleed Side - Plan and Elevation ... A-56
- 6. Electrical Drawing Notes and Symbols . . . . . . . . . . . . . . . . . . . . A-62:
- 7. Electrical One-Line Diagram ~......................~....... A-63
- 8. Control Logic Notes and Symbols .................'........ A-65
- 9. Control Logic Diagram - F&B Initiation Logic and Bleed Valve Control ..................................... A-67
- 10. Structural Details - New Pump Foundation - Feed Side .... A-72
- 11. Structural Detail - Platforms at Elev. 78'-2" &
63'-6" - Bleed Side ..................................... A-73
- 12. Implementation Schedule ................................. A-85
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LIST OF TABLES Page
. 1. Feed & Bleed Cases Considered ........................... A-32 1A. Feed & Bleed Cases Considered (Alternate) ............... A-33
- 2. Feed & Bleed Flow Requirements .......................... A-34 2A. Feed & Bleed Requirements (Alternate) ................... A-35
- 3. Pump Data ............................................... A-36 l
- 4. Cost and Connected Load Comparison for New Pump for Feed and Bleed Service .................................. A-37
- 5. Bleed Valve Data ........................................ A-38 5A. Bleed Valve Data (Alternate Cases) ...................... A-39
- 6. Classifications for Equipment and Piping ................ A-41
- 7. Feed & Bleed System Components .......................... A-44
- 8. New Pump Summary Specifications ......................... A-49
- 9. Bleed Valves ............................................ A-50
- 10. F&B Quench Tank ......................................... A-53
- 11. Electrical Loads Required for Feed & Bleed Operation .... A-59
- 12. Cost Estimate Summary ................................... A-80 t -
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A-5
LIST OF SYMBOLS .
'ADHR -
Alternate Decay Heat Removal AFW -
American-Nuclear Society ANSI -
American National Standards Institute APS -
Arizona Public Service Co.
ASME -
American Society of Mechanical. Engineers BFP -
Brake Horsepower CD -
Controlled Depressurization C-E -
Combustion Engineering Co. .
ECCS -
Emergency Core Cooling Systems F&B -
Feed and Bleed FSAR -
Final Safety Analysis Report HP --
.High Pressure .
Hp -
Horsepower. ,
, c HPCS -
HPSI -
High Pressure Safe'cy Injection ISI -
In-Service Inspection kW' -
tilowatts LOCA'
- css of Coolant' Accident-MW -
Megawatts MWe --
.' Megawatts Electrical MWt' '- Megawat'ts Thermal' NPSH -~ Net-Positive 1 Suction Head '
.NSSS -
Nuclear Steam' Supply System l
~A-6
SYMBOLS (Cont'd.) ,
PORV -
Power Operated Relief Valve PRZ -
Pressurizer PVNGS -
Palo Verde Nuclear Generating Station PWR -
Pressurized Water Reactor RC -
Regulatory Guide RHR -
Residual Heat Removal l RWST -
Refueling Water Storage Tank SCE -
Southern California Edison Co.
SDC -
Shutdown Cooling SDG&E -
San Diego Gas & Electric Company SIS -
Safety Injection System SLC -
Standby Liquid Cooling SNL -
Sandia National Laboratories SRV -
San Onofre Nuclear Generating Station TLOFW -
Total _ Loss of Feedwater USNRC -
U. S. Nuclear Regulatory. Commission
.9 A-7
1.0 INTRODUCTION
~
This study has been performed to evaluate the feasi-bility,. costs and effects on operation of installing a system in presurized water reactors (PWRs) that would provide the .
capability, under certain accident conditions, to remove decay heat loads at hot standby, and to cool down and depres-surize the reactor to achieve and maintain cold shut-down.
Emergency core cooling systems in present PWRs are designed to inject sufficient water to cool the reactor core if the primary system should undergo a rapid depressurization as the result of the classic large loss-of-coolant accident (LOCA).
Accordingly, high pressure safety injection (HPSI) pumps are frequently designed to start injecting water at some pressure (e . g . , 1600 psia) well below the normal operating pressure of the primary system. Therefore, a number of present emergency cooling systems do not have the capability to inject water at or near normal operating pressure in an accident situation where there has been no large loss of coolant that could have caused a rapid depressurization.
Concern has been raised about accident conditions under which the ability to remove heat through the steam generators has been lost while the primary pressure remains high.
Under such conditions the pressure would build up as the result of decay heat addition to the coolant until the primary safety relief valves (SRVs) lift and release coolant to hold the system pressure at this level. The continued loss of coolant through the SRVs with no accompanying makeup, 3
other than a small, inadequate amount from the charging pumps, would eventually lead to uncovering of the core.
During all of this time the HPSI pumps would not have been abla to inject water because the primary pressure remained at the elevated level. 'One distinct accident condition that could lead to a complete loss of heat removal capability through the steam generators is a total loss of feedwater (TLOFW), such as would occur if all normal and emergency j feedwater pumps became inoperative. Another is steam generator tube leakage of sufficient magnitude as to require isolation of the steam generator but large enough to depressurize the primary system and allow HPSI flow to be initiated. The con-corn for this condition is less than that for a TLOFW since all steam generators would have to be affected before a total loss of decay heat removal would occur.
A potential solution for this situation is a high pressure .
" feed'& bleed" (F&B) system with the capabilities to do the following:
- 1) Inject sufficient water to remove decay heat and prevent core uncovering at system pressures A-8 t
~ ,
l which are at somewhere between the normal operating l
- value and the lifting pressure of the SRVs. This corresponds to maintaining the reactor in the hot standby condition.
l 2) Cool down and depressurize the primary system to the pressure level where the HPSI or low pressure L safety injection (LPSI) pumps can be used to perform the long-term cooling function and maintain cold shutdown.
l Another solution is a system which bleeds coolant rapidly enough for the pressure to drop and the existing HPSI pumps'to
! inject water in a timely manner.
The San Onofre' Nuclear Generating Station, Unit 2, (SONGS)
! of Southern California Edison (SCE) and the San Diego Gas &
Electric (SDG&E) Company was selected as a. prototypical PWR for investigation of the feasibility of installing a F&B' system on
- a. plant which does not have that capability. SONGS is a 3410 i megawatt thermal.(MWt):PWR plant with the Nuclear Steam ~ Supply System -(NSSS) provided by Combustion Engineering (C-E). It is ,
e in the final stages of construction. After development of a -
conceptual system design, investigation of the feasibility of-its installation, and estimates of its costs for SONGS, a i.
brief assessment was made of factors which could= affect feasi-bility and costs of installing the same type offsystem in other plants such as the Palo Verde Nuclear Generating 1 Station (PVNGS) of the Arizona-Public Service'(APS): Company.. This plant has a C-E 3800 MWt reactor, the later. System 80 design, and is in a'somewhat earlier' stage of construction than SONGS.
! 1 Prior to performance of this study, a TLOFW transient l
analysis was performed by C-E (Ref.1) someL of the results of which have been used in establishingLthe basis for this study, including the following (Later analyses yielded more optimistic results) :
- 1) .During the 10 minute interval.~followingLTLOFW "
boiling; dry of the secondary. side of the steam-
. generator; occurs. This is based on'the availa -
bility of off-site power, a reactor trip after
! 30 seconds and tripping.of'the primary coolant' pumps after'10 minutes.! Initially.there.is a decrease in pressure because"the heat removal capability.of the steam generators is greater.than-
~
the decay heat load. As the boil-off7of the secondary progresses,'the. heat removal _ capability o dr' ops below thel decay heat' load,-causing the-primary, system pressure'to rise until~the SRV set
~
c pressure of~2500 psia:is reached.- ; As the TLOFW:
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continues, the SRVs open and close as necessary to "
to keep the pressure at 2500 psia. This results in a major net loss of primary coolant as the only makeup is now from the charging pumps which have an ~
injection rate much lower than the flow rate through the SRVs.
- 2) With TLOFW (TLOFW implies no AFW) and no corrective action, such as F&B, the two phase mixture level in the reactor inner vessel (core area) drops to the core mid-plane level in approximately 55 minutes.
- 3) A F&B operation starting at approximately 20 minutes after reactor trip was analyzed. In this case bleed was by means of two new power-operated relief valves (PORVs), which remain open after being actuated in order to bring pressure down to-where one of the existing HPSI pumps could start injecting water at approximately 1400 psia. This resulted in the two-phase mixture level dropping approximately one-third of the core length below the top of the core before starting to rise again.
- 4) A third transient analysis showed that the latest time after reactor trip for which AFW could be re-stored without resulting in core uncovery was 50 min-utes after the reactor trip.on loss of feedwater..
Note that'in these analyses, whose results are summarized above, the level of the two-phase mixture is referred to rather than the liquid coolant level. (collapsed water level) . The exact nature of the two-phase mixture,-its variation with
-height and its effectiveness in preventing fuel damage -(in comparison to all liquid) were not described in the material made'available for this study.
By directive of Sandia National Laboratories (SNL), two basic cases were investigated:
l 1) Case 1 - The addition of two new PORVs, sized in accordance with the C-E transient analysis with water injection being provided'by one of the 1 existing HPSI pumps.
l 2) Case 2 - The addition of two new bleed valves in com-
! .bination with upgraded HPSI' pumping' capability to ,
l provide a F&B capability'which would prevent or minimize
- uncovering of-the core. It was stated by SNL that-this case was to be investigated on-the basis of.HPSI -
initiation at 50 minutes after reactor trip *and A-10'
with the flow rate necessary to remove decay heat
, at that time by water heatup only (solid water system).
This time was used to match the time at which the C-E results showed that restoration of AFW would
. still prevent core uncovery.
Preliminary analysis, using basic engineering consider-ations rather than a RELAP type of computer analysis, indicated the possibility that the F&B operation described above in Case 2 would not prevent uncovering of the core; therefore, two addi-tional F&B cases were investigated..
o Early initiation of F&B at 10 minutes, which corres-ponds to the boilout time of the steam generators, at a rate sufficient to prevent core uncovering o Delayed initiation of F&B at 40 minutes at a rate sufficient to prevent core uncovering. This is the latest time that F&B can be initiated with a reason-able assurance that the core will not be uncovered.
In the case of delayed initiation of F&B, an alternate possibility is that of opening the bleed valves early.to-prevent lifting of the SRVs. For the two cases with delayed pump actuation at 50 and 40 minutes, an alternate system design in which the bleed valves would open early at approximately 10 minutes at a pressure below the set point of the SRVs was also considered.
The overall study was carried out in sequential parts which are outlined below and reported in detail in the cor-responding sections:
Section 3.0 - Various F&B methods were considered in terms of when to initiate F&B and what l'njection flow-rates to use; whether to upgrade, replace or supplement ~
the existing HPSI pumps; what type of bleed valves to use; and how to handle steam discharged from the bleed valves.
A method was selected and defined as a base for completion r of'the study.
Section 4.0 - A conceptual design was prepared for the-selected method, including design criteria,-definition of the basic F&B system, description of major components, and_the supporting electrical, instrumentation & control
_, and structural features of the design.
Section 5.0 - Implementation of the' selected method was investigated with respect to the installation requirements for. equipment, piping and wiring;-testing of the system; costs and scheduleffor designing, purchasing, installing and placing into operation; and the resultant radiation-exposure.
A-11
Section 6.0 - The operation and maintenance requirements for the selected F&B system were examined.
Section 7.0 - The requirements for the detailed engineer- '
ing, design and analysis required to proceed with the conceptual design to the point where implementation of a F&B system in an existing plant or one under construc-tion-could begin are defined.
Section 8.0 - An assessment is made of the applicability of the method selected and conceptual design established for installation in PWR plants other than the specific.
plant considered in the study (SONGS). 'The objective is to determine the broader applicability of the study results and wnere major differences might be expected from plant to plant.
The results of this study'are summarized in the Executive Summary of'Section 2.0.
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l 2.0 EXECUTIVE
SUMMARY
Purpose of Study This study is part of the program for the U. S. Nuclear Regulatory Commission (USNRC) Unresolved Safety Issue (USI)
A-45 dealing with reactor decay neat removal. It was per-formed to investigate the feasibility and costs of implementing controlled depressurization of the primary system in pressur-ized water reactors (PWRs) using power operated relief valves (PORVs) or other types of bleed valves. Such a capability may be required for certain accident conditions such as total loss of feedwater flow (TLOFW) where there is no rapid loss of primary coolant that would result in rapid depressurization to the level at which the high-pressure. safety injection flow is initiated.
Study Methods As defined by Sandia National Laboratories (SNL) in the required scope of work for the study, two cases were inves-tigated. The basic case considers adding two large PORVs, along with associated piping and equipment. These PORVs have relieving capacity well in excesslof that required for decay heat removal in order to depressurize the system as rapidly as possible to permit an existing HPSI pump to initiate flow injection.
Valve capacity and time of opening after TLOFW for this case were based on the results of earlier transient analyses performed by Combustion Engineering (C-E) (Ref. 1).
The second case investigated considers upgrading both feed and bleed capabilities by means of supplementing the ex-isting HPSI pumps to permit flow injection to be initiated at or near full system pressure'in addition to adding bleed valves. This is a more conservative approach than that-of the first case where an-existing HPSI pump is relied upon. Investi-gation of this second case was based on two factors:
- 1) Uncertainties in the final definition of the criteria for protection against core uncovering;-i.e., whether the require-ment is that the top of the two-phase mixture zone remain above the top of the core, some brief uncovering is allowed or the more coiservative requirement that the-collapsed water level
~
must remain above the top of the core. The last would require
. an upgraded HPSI capability.-
- 2) The uncertainties about the~modeling used forftransient analyses and interpretation-of the~resultsLobtained.
I In considering upgrading of the HPSI pumping capability the' required flow rate to be; considered, as specified by SNL,
.A-13
l l
was that necessary to remove decay heat by water heatup only; ^
and at the latest time after TLOFW for which the C-E transient analysis showed that feedwater could be rastored without core uncovering. Approximate calculations, which included ,
allowance for the transit time for the cool injected water to reach the core, indicated that this injection fic: rate would lead to core uncovering as defined by the collapsed water level. Therefore, two additional high injection flow rate cases at above normal primary system pressure were con-sidered which would prevent the collapsed water level from dropping below the top of the core:
- 1) Early actuation of the new pump at a few minutes after TLOFW.
- 2) Accuation of the new pump at the latest possible time after TLOFW which will maintain the equi-valent liquid coolant level above the core.
The latter of the above two cases was used to develop a conceptual design based on the most stringent requirements for pump flow and electric power. It is expected that ultimate selection of pump size based on more definitive core uncovering criteria and more detailed transient analyses would result in a flow requirement lying below the bounding value used. A combin-ation of full opening of the PORVs to provide rapid depressuri-zation and the use of a new pump having head and flow character-istics intermediate to those used in the bounding case of this study and those of the existing HPSI punp may be shown by further transient analysis to be capable of proveating core uncovering on the more conservative basis of collapsed water level.
Transient analyses made by C-E in early 1983 (Ref. 1) have a relationship to this study in the following respects:
o The total relieving capacity of the PORVs here is the same as that used by C-E.
l o Water level information in the TLOFW transient analysis was used as an input condition to deter-i mine upgraded HPSI flow requirements to prevent
! core uncovering on a collapsed water level basis.
l o With new large PORVs and an existing HPSI pump the vessel two-phase mixture level drops to about .
one-third below the top of the core, which tends to substantiate consideration of an upgraded HPSI l pumping capability as an alternative to depending -
l on an existing pump.
A-14
g o The C-E transient analysis was based on no' loss of off-site power with the reactor being tripped at 30 seconds and the reactor coolant pumps at 10 minutes
, after TLOFW. This study also took into account the possibility of a loss of off-site power during a TLOFW condition and the requirement that the feed and bleed system would have to be powered from an amergency diesel generator.
Results of transient analyses performed by Argonne National Laboratories (ANL) for conditions similar to those of the C-E analysis indicate that the two-phase mixture level may not drop below the top of t.te core. These results indicate that s TLOFW with no loss of_off 9ite power may not require an up-graded HPSI pumping capability; and that TLOFW combined with loss of off-site power is even less likely to equire this.
Thus, the requirement for an emergency diesel generator to supply a new and larger pump becomes less likely.
For each of the two cases considered two subcases were investigated:
- 1) Installation in a new plant during the final stages of its construction
- 2) Installation in a plant that had been operating for some time.
The San Onofre Nuclear Generating Station (SONGS) was selected as a plant not currently having a depressurization capability to determine the feasibility and costs of imple-menting such a capability.- At the same time the applicability of implementing the upgrading in-other plants of similar design was examined to determine what aspects of the' design could make a significant difference on a plant-specific basis.
Conceptual Design for Plant Upgrading Referring to the discussions in the previous section.on
" Study-Methods" a conceptual design was established for a com-plete feed and bleed' system embodying the following features:
o Capability of being retrofitted in SONGS a o A bleed' capability consistent with that provided by;large PORVs in the C-E transient analysis
~
o AL feed capability which would assure.no core uncovering,-based on-the collapsed' water level, at a sufficiently long time after TLOFW to' ascertain
-if-auxiliary'feedwater flow could be restored.
A-15~
The conceptual design for the complete new feed and bleed (Case 2) system also covers the case where new bleed valves ~
are used in combination with an existing HPSI pump (Case 1) ,
except PORVs rather than control valves are used with the .
existing pump.
A full set of design criteria has been developed for the system design. The more important of these criteria include the requirement that the system equipment and piping must be consistent with the existing safety injection systems with respect to ASME - Code Class, Nuclear Safety Class, Quality Group Class and Seismic Category; the new valves must be fully safety grade and environmentally qualified; the system must be capable of operation when off-site and turbine-generator power sources are unavailable (e.g., from a single existing diesel-generator); the new system must in no way affect the functions of the existing safety injection systems; and redudancy is not required (e.g., the single-failure critierion does not apply).
Bleed side mechanical design consists of the two bleed valves mounted at the pressurizer top using the nozzles pro-vided for the existing safety relief valves (SRVs); a quench tank (similar to the existing quench tank) and connecting piping. In the event of an extended TLOFW condition, the quench tat.k will fill and its rupture disk will permit the blowdown from bleed valves to_. flow to the containment sump and eventually be used in the shutdown coolin? (SDC) recircu-lation mode. The new quench tank would only serve to contain minor, occasional releases from the bleed valves, similar to the way the existing quench tank serves the SRVs.
Feed side mechanical design includes the new pump with its suction taken from the refueling water storage tanks (RWSTs) and its discharge connected to the four low pressure safety injection (LPSI) lines to the reactor coolant loops.
These connections are made at a point where the existing lines have changed design pressure rating to the high pressure level of the reactor coolant system but outside of primary contain-ment. No new containment piping penetrations would be required.
Consideration was given to possible upgrading of an existing HPSI pump or the addition of a booster pump. Both of these were found not to be feasible; therefore, a co.apletely separate pump capable of injecting at or above full reactor coolant system '
pressure has been specified. Room was found for this pump in the SONGS compartment containing the No. 2 HPSI and LPSI pumps; A-16
.. - .. ~ . -- - - . - ..
k however, neither of the other two pump compartments would have room for a redundant pump if the single-failure cri-terion were imposed.
The main consideration in the electrical design is the
( ability to supply power to the new, large pump from an existing
- diesel-generator. If neither the existing HPSI or LPSI pumps would be required or allowed to run while the new HPSI pump was in operation, operation would be possible with one SONGS diesel-generator but.with only a small reserve margin with respect to the machine's nameplate rating. Supply of electri-cal power for the first case where an existing HPSI pump is used poses no problem.
1 New 4160-volt switchgear, circuit breakers and a 480-volt i motor control center would be required.
F No major structural changes or additions would be required to accommodate the. feed and bleed system. Structural work would consist mainly of additional pipe supports, pump base, '
platforms, walkways and railings.
! In the present conceptual design the basic control for Cases 1 and 2 is automatic; thatfis, the valves would be opened l'
and the new pump started in response to a higher than normal primary system pressure and a low feedwater flow signal. In-case 1 the PORVs would be fully opened to drop the pressure i as rapidly as possible to the level where an existing pump
. would initiate flow. In the full feed and bleed, Case 2, the
-bleed valves would open and the new pump start in response to the same high pressure and low feedwater flow signals. How-ever, the bleed valves would function in a modulating mode to
! hold pressure constant until the time when decay heat gener-ation is low enough (about 3 hours) to permit the shutdown i
cooling mode (SDC) of the LPSI pumps to take over. At this time the bleed valves would be fully opened and the new pump stopped.
In the ultimate detailed design of a control system, consideration may be given to an all-manual control system
! for the following reasons:
1 o Simplicity of control and avoidance of spurious initiation o Elimination of the need to interface with existing
- primary pressure and feedwater flow instrumentation channels; thus; no possibility of-jeopardizing i.
these channels -
o Lower implementation. time and. costs A-17
r o Operator choice to maintain full primary pressure to the latest possible time after TLOFW for possible restoration of auxiliary feedwater for cooling.
Full instrumentation for flows, pressures, tempera'ures c
and levels are included in the design. Special instrumentation is included to sense accidental opening of the valves. This includes measurement of different parameters and the use of state-of-the-art instrumentation developed as the result of TMI 2 experience.
Detailed Design, Installation and Operation of a System for Plant Upgrading The detailed engineering and design, based on the concep-tual design developed in this report, would be of the type normally performed for nuclear power plant safety systems. Be-cause of the expectation that a system for a particular plant would either be designed and installed during the latter stages of overall plant construction, or retrofitted to an operating plant, the engineering and design would have to be organized as a separate project with a dedicated project team.
The nature of the system application will require a sig-nificant amount of special analysis. Before actually proceeding with the design, it would be necessary to perform additional transient analyses to determine: (1) if the addition of-new PORVs in combination with an existing HPSI pump would be adequate to meet the established criteria (e.g., core uncovering and/or fuel i clad temperature); (2) and, if not, what injection flow rate and new pump size would be required. Electrical analysis to establish which loads must operate during feed and bleed, the ability of the existing diesel-generator to carry these, and
~
the effect of starting the large new pump in the automatic loading sequence.
Analysis of radiation levels during plant shutdown is a requirement for installation planning. Because of.the~ safety classification of system piping.and equipment major stress analysis and pipe break analysis efforts will be required.
High thrust loads'from bleed valve blowdown and-stress analysis in " super-pipe" locations; (e.g. , at- the point of attachment to existing-safety injection lines) will require particular attention. ,
A considerable analytical effort'will be required.to ,
select the actual valve type-for-the bleed application. The analysis will have to cover the ability to handle all steam, two-phase mixture and eventually solid. water; the ability to meet safety grade and environmental requirements; and thelre-A-18
quirements for pressure relief vs. modulating control type
. of operation.
Other analyses to be performed will include the effects
+ of a solid water mode of primary system operation, cold repres-surization, and primary system cooldown rates.
. Installation of new PORVs or a complete feed and bleed system will have to be carefully planned and executed because of the fact that essentially all plant equipment and piping will be in place at the time it is done. In the cases of
~
an operhting plant, there will be the added complication of avoiding interference with plant operations and safety, as well as avoiding radiation exposure to personnel in certain loca-tions. In.these cases as much of the work as possible would have to be done while the plant is operating in order to minimize that to be done during scheduled outage. Radiation levels in areas such as the pump room, switchgear locations, control room, and other areas outside of containment should not impose serious restrictions on working crews during plant operation.
Coordinated schedules for engineering, design and analysis and construction have been developed. The former has a span of 18 months and the latter 12 months. There is a 6 - month cverlap resulting in an overall project schedule of 24 months. The schedule is keyed to an annual outage for refueling and seneduled maintenance, which is assumed to be of 2 months duration. Start of project work would be dependent on a sufficient base of NRC.
rulemaking and criteria development having been established. The major critical path item of the schedule would~be the new pump procurement, which-is expected to require 13 - 14 months.
Installation of the system within the planned time would depend.
to a large extent on careful planning for doing as much of the work as possible while the plant is in operation in order to
. minimize'the work to be done during the scheduled outage.
In retrofitting a feed and bleed system to a plant that has been in operation for some time, radiation exposure to personnel will be a concern. The two potential problem areas involved are around the pressurizer within containment and adjacent to the LPSI pump in the pump-cubicle. Radiation levels can build'up at-the latter because of the use of the LPSI pump for shutdown cooling. Data from several plants that have been operating for a number of years. indicate that shutdown radi-
~
ation levels can be as high as 0.3 - 0.4 R/hr at certain
- . specific locations in these two areas.- Such levels would sev-erely limit the time that personnel could spend in the area-
.- during' installation. In the case'of the pump area, the use of
- temporary shielding should' attenuate radiation levels sufficiently
! to. eliminate any need1to limit occupancy time. .Since the use~of temporary. shielding.in the pressurizer area may not suffice, it-has been. conservatively assumed that allowance would be required' for burnout of' installation personnel.
A-19 i
Although the feasibility of installing a system for pro-viding feed and bleed capability was investigated specifically .
for SONGS, the conceptual design and other material developed in this report would have a generic applicability to other plants lacking this capability. Whether such an installation -
could be technically feasible and possible within reasonable cost limitations would be highly plant specific and such a determination would require an individual detailed investi-gation.
Factors that could be expected to affect the feasibility for a specific plant would include:
o The arrangement of equipment and piping around the pressurizer and the availability of a suitable connection for the bleed valves o Room for the installation of a new pump if required o Accessibility of safety injection piping for the necessary interconnections o Capacity of the existi.ng emergency diesel generators to carry the additional electrical load.
Once a bleed valve type has been selected and qualified for the application in the first plant, it should be suitable for use in any subsequent plants being upgraded. This is also true of a new pump, if found to be required.
Implementation Costs Estimates to cover the complete cost of implementation for the two basic cases, addition of PORVs with an existing HPSI pump and addition of bleed valves with a new pump, were made for installation in a new plant under construction and a plant that had been in operation for some time. The total estimates for each of these cases are shown below:
o Case 1A - PORVs only, new plant - $2,495,000 o Cast 1B - PORVs only, operating plant - $4,254,000 o Case 2A - Bleed valves and pump, new plant -
S6,958,000 =
o Case 2B - Bleed valves and pump, ,
operating plant - $10,310,000 A-20
Construction costs and costs for supporting services were estimated separately. Construction costs were subdivided into
- mechanical equipment and piping, structural, electrical and instrumentation and control work. Included under supporting services were project management, engineering design and
~
analysis, quality assurance, construction management, testing and startup, training, and costs related to health physics and radiation exposure control.
Prevailing construction labor rates in the San Diego area were used and allowances made for 3 - shift operation, premium time on weekends, overtime at shift changes for work during the scheduled plant outage, and travel allowance for construction workers. In the case of installation in an operating plant, allowance was made for the additional manhours and other costs associated with burnout of craft labor personnel in high radiation areas and also for the general difficulties associated with working in an operating plant.
' Present day costs were used and escalation applied at 6%/ year using the developed schedule. Allowance was made for interest during construction at an annual rate of 12%. An overall contingency allowance of 25% was made.
There are two cost factors not included in the present estimates which could have a very significant effect in another plant specific situation:
- 1. If the addition of new pump is required, the additional electrical load may not permit use of an existing emergency diesel generator. Costs associated with the addition of a dedicated diesel-generator for the feed and bleed system, including structures and supporting systems, could be roughly
$11,000,000, which is somewhat greater than the total of all other implementation costs. This would not be a concern where PORVs only'would be added.
- 2. In the case of an operating plant, replacement energy costs incurred by the prolonging of a scheduled' annual outage by the installation of the system could result in costs that would exceed the total of all other-implementation costs in just a few days, considering that replacement energy costs are typically in the range of $500,000 to S1,000,000 per day. In an l
actual installation.if the work could not be all completed in the period of one annual outage, it
~ could be completed during the following year's l
- outage, if the NRC rules permitted. The necessity for hydro-testing at the completion of system instal -
lation may extend the outage by two or three days; however, as is often the case, the turbine-generator A-21
may be on the critical path in determining the outage time.
Controls for.the full feed and bleed system were estimated on the basis of automatic control. The alternative use of manual controls would reduce the costs by approximately $230,000; however, costs would not be a primary consideration in selecting automatic versus manual control.
Conclusions The following general conclusions have been reached as the result of the study performed:
, o For PWR plants lacking the capability, addition of a system to permit controlled depressurization would be feasible. Such a system could consist of PORVs which would permit rapid depressurization to the level where an existing HPSI pump could begin flow injection; or, it could consist of bleed valves and a pump added to provide a feed and bleed'capa-bility.
o Transient analyses performed by others' tentatively
. indicate that the use of PORVs with an existing HPSI pump may prevent core uncovering. However, a final decision on whether an upgraded pumping capability will be required can not be made until firm criteria for fuel protection have been established in terms of amount and type (two-phase mixture vs. collapsed water level) of core uncovering and/or maximum fuel ~
clad temperature-allowable. Further transient analyses may be necessary to reach.this~ decision and to establish the required _ injection flow-rate if upgrading of pumping capability is found to be ne-cessary.
o For purposes of_ developing a conceptual' design and investigating the_ feasibility of implementation, an-upper bound case of injection ~ flow rate and'asso-ciated pump: size has been determined lon the conserv-ative basis.of no core uncovering in terms of collapsed
. water level and the requirement to-provide injection above normal system pressure.. Depending on the results c of-further transient analysis, it may be possible to meet the. conservative criterion for no core uncovering.by use_of a new pump having head and flow requirements-less than those of the pump speci-fled in'this-study but in' excess of the existing HPSI.
pump capability.
A-22
o A conceptual design that is applicable to either the full feed and bleed system or to the use of PORVs only has been developed. This could serve as the starting point for the proceeding with
. detailed engineering and design for a specific plant.
o Detailed engineering and design for a specific plant would have to be supported by a significant amount of specialized analysis, o Installation of a system would have to be very carefully planned and executed, particularly in an operating plant.
o An overall schedule of two years from start of engin-eering and design to completion of installation and testing is considered feasible. For an operating plant, keying of schedule to an annual scheduled out-age would be essential.
o Radiation exposure to personnel for installation in an operating plant will have to be taken into account but appropriate allowances can be made, o The conceptual design developed is generally applic-able to other PWR plants requiring the depressuri-zation capability. However, certain factors such as diesel-generator capacity, equipment layout and piping arrangement will strongly influence the feasibility and costs for a given plant.
o Implementation costs range from $2.5 million for PORVs only in a new plant to $10 million for a complete feed and bleed system in an operating plant. The need for a dedicated diesel-generator or the incurring of replacement energy costs for extended outages could increase these estimates very significantly.
O e
A-23
3.0 FEED & BLEED METHODS FOR CONTROLLED DEPRESSURIZATION .
3.1 Methods Considered In order to select an F&B system for development into a conceptual design, the following important characteristics and the possible variations were investigated:
o The time after reactor trip that F&B is initiated and the associated injection flow rate.
o How the steam released from the bleed valves is to be quenched; e.g., by release to the open containment or by condensing in a closed system.
o Means of making the transition from primary system pressure at or above the normal system pressure to the reduced pressure at which the LPSI pumps in the SDC mode take over, o Upgrading of the existing HPSI pumping capability by means of upgrading an existing pump, replacing an existing pump, or by installing a new pump that would separately provide the desired feed capa-bility, leaving the existing HPSI pumps as is to perform their function to inject water under LOCA conditions.
o The type of valves appropriate to the required bleed function. For example, there is a choice between a true power-operated relief valve which is either fully open or closed in response to an external pressure signal or a modulating type of control valve that is positioned in accordance to a pressure error cignal.
3.1.1 F&B Initiation Time and Injection Flow Rate Establishment of the F&B initiation time and the injection flow rate is the primary and most basic factor in establishing the F&B method to be used. The initiation time, which is the time elapsed between the reactor trip resulting from the accident condition (e.g., TLOFW) and the initiation point, can range from that early time when the-steam generators boil dry and the primary pressure approaches the set point of the SRVs to the latest time at which AFW -
can be restored without uncovering of the reactor core.
Further, feed and bleed need not necessarily be initiated at ,
the same time. Possible injection flow rates vary from the upper value required to remove decay heat by boil-off only to the lower value set by water heat-up and boil-off.
A
, The basic need for an F&B system arises from the fact that typically PWRs do not have the capability to inject water at or near system operating pressures in quan-
. tities sufficiently large to remove decay heat and prevent core uncovering. Since HPSI systems are typically designed for a major LOCA in which pressure is reduced rapidly to a lower pressure at which injection can take place, there is no provision to inject cooling. water in an accident situ-ation not-involving a large LOCA, such as a TLOFW, other than by charging pumps which-generally have a flow rate less than required for decay heat removal.
As noted in Section 1, at the start of this study, two basic cases were identified for investigation:
Case 1 - Addition of new PORVs which would have an excess. relieving capacity (e . g . , in excess of that
, required for decay heat removal) in order to reduce L pressure and permit flow injection from an existing l HPSI-pump.
Case 2'- Addition of new bleed' valves in combination with an upgraded HPSI pumping capability.that would permit injection'at-full' system pressure.
In-the discussions ~whichLfollow, reference is'made-to Tables 1 & 1A which show the definition of the cases considered, Tables-2 & 2A which show the required F&B' flow rates.for pumps and valves, and Tables 3,'4 and 5 & SA which show pump and valve data-for the different cases. These tables are at the-end of-this section.
Conditions for Case 1 are established directly by-the information appearing in the'C-E transient analysis (Ref. 1)..'In this case:therelis no loss of off-site: power,;
the reactor _is tripped ~atL30 seconds and the reactor coolant pumps at 10 minutes after~TLOFW.- The steam' generators boil dry and the pressure rises:to 2500 psia'at 10 minutes after TLOFW At this point'the'SRVs'open and. relieve-steam to' hold pressure at 2500 psia.until F&B'is initiated at=20~ minutes ~
-(presumably-non-automatically). Because of.the-propose'd:
excess relieving capacity of:the:PORVs over that requiredf ifor_ decay heat' removal, pressure decreasas until;at approx-imately 45 minutes'the. existing HPSI pump starts;to inject:
water;and at approximately 55l minutes the pump flow reaches-e- an equilibrium-with.the' pressure / flow characteristic of'the-two PORVs.
.Ref. 1 shows ' that the - two-phase. mixture: level -
.,. dips to'a: point approximately;one-third the core length
-below:the:topfof the core beforera recovery. starts. :The results-showing,the'levelJof the two phase mixture:do:not indicate-its-void' fraction orJits; effectiveness.in: cooling the; fuel. This 4
~A-25:
1
raises a concern as to whether or not there would be adequate heat transfer capacity to prevent core damage, even if the .
two-phase mixture level had not dropped below the top of the core.
Case 2A, which includes upgraded pumping capability, calls for bleed valve and pump actuation at 50 minutes, after TLOFW which Ref. 1 shows to be the latest time for which AFW can be restored without the two-phase mixture level going below the top of the core. The injection flow rate is stipulated as that negessary to remove decay heat by water heatup only (from 120 F to saturation temperature at 2500 psia).
This contemplates an eventual solid-water system through the reactor vessel and pressurizer and out the valves. The problem anticipated here is, that in the transit time for the injected cool water to reach the core mid-plane (approx-imate calculations indicate about 30 to 45 minutes), the boil-off to remove decay heat will remove sufficient water to bring the level well below the top of the core before decay heat removal by water heatup becomes the dominant cooling mechanism. This results from the fact that h is only about 60% A h The transit time is that required k8 displace the hot water ib. the reactor vessel annulus, the bottom spherical head space and halfway up the core. Credit is not taken for heat-up of water already in the primary pipes and reactor vessel, which is a conservative approach.
All calculations involving decay heat generation rate were made using ANSI (ANS) 5.1-1979 (Ref. 4) and assuming a rate 20% greater than the median curve.
Because of the concerns raised for Case 2A, two alternate cases were investigated:
Case 2B - An early initiation of F&B at about 10 minutes after TLOFW at the time the steam generators boil dry permits a lower flow rate than delayed initiation by taking credit for boil-off of water above the top of core while cool water is in transit to the core. This procedure would require a commitment to at least have F&B before the time at which AFW, if restored, could prevent core damage. However, if initiated, it could be terminated if AFW were subsequently restored. Boil-out of the steam generators at 10 minutes is based on the availability of off-site power with a reactor trip at 30 seconds and primary coolant pumps tripped at 10 minutes into the transient. An earlier trip of the -
reactor and pumps would delay steam generator boil-out.
Case 2C - This case which is based on that time at which l
all of the water inventory above the top of the core has been boiled off and water must be injected at a rate A-26
corresponding to decay heat removal by boil-off only to prevent core uncovering, would provide a reasonable time to- determine if - AFW cculd be restored (40 minutes versus 50 minutes for Case 2A) but would require the largest flow rate of the three cases.
Another decision to be made in F&B initiation is whether the bleed valves would be opened when the pump is started or earlier (e.g. , at 10 minutes) . The advantage in opening the valves at 10 minutes would be avoidance of lifting of the SRVs. In either case, the amount of fluid discharged through the SRVs or bleed valves, until injection starts, would be the same. Cases for early bleed initiation as an alternative have been identified as 2A-A, 2B-A, and 2C-A.
The next factor to be considered is how the steam discharged from the bleed valves is to be quenched. Three methods were considered:
- 1. Discharging into the existing pressurizer quench tank which takes the discharge from the SRVs.
- 2. Discharging from the new valves into a new quench tank having the same design characteristics as the existing quench tank.
- 3. Discharging into a closed system that would con-dense the released steam and return water to the primary system.
It has to be recognized.that discharging into a quench tank, whether new or existing, would not prevent the ultimate release of the discharged fluid to the containment.
During the F&B operation the quench tank rupture disk would open and release water to the containment sump after the tank had been filled since the amount of water released from the primary system would considerably exceed the capacity of the tank.
Brief consideration was given to the use of a closed system that:would condense the steam released from the PORVs and return it to the primary system without release to containment. This method was eliminated because of the large-heat load involved, the size andEcomplexity-of the equipment, and the impact its. installation would have on a
, PWR plant;-e.g.,'very probably an extension of containment.
Further, it would be very1 difficult-to justify a system with a low probability of ever being.used.
In keeping with thel requirement to disturb exist-ing' plant' equipment as little.as possible, it was' decided to A-27L
install a new quench tank, identical to the existing tank, with piping to the new PORVs. As in the case of the exist-ing SRVs and quench tank, this tank would only serve to -
contain fluid released during a pressure transient, such as that from a full-load rejection, but could not contain the major release during an extended TLOFW situation. The tank
- would just act as a pass-through device to the containment sump after its rupture disk failed.
The next decision point concerns how the F&G system will be used to depressurize from the initiation level of about 2400 psia to the pressure at which the LPSI pumps in the SDC mode
- can take over the decay heat removal function'.
This latter pressure is 150 psia and normal SDC occurs approxi-mately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after reactor trip. Two choices are available:
- 1. Use of a programmed reduction of bleed valve set point to follow the decay heat curve.
- 2. Maintain pressure at 2400 psia until ready for the LPSI pumps to take over in the SDC mode and then rapidly reduce pressure by stopping the new pump and opening the valves fully.
Both of the above methods would require valves that could be modulated, as well as fully opened or closed.
The second method was selected for two reasons:
- 1. Better assurance of an adequate margin of sub-cooling to prevent boiling in the core.
- 2. With a solid-water system, which this will be after some initial period of steam blowdown and two-phase flow through the valves, system pressure will be very sensitive to control changes in the position of the valves. Maintaining a continuous high pressure will reduce the pressure swings due to control action.
The HPSI pumping capability could be upgraded to meet the higher pressure and flow conditions by one of the following methods:
- 1. Upgrade an existing pump by adding a booster pump to provide the additional required head.
- 2. Replace one of the existing HPSI pumps with a new HPSI pump having the higher head and flow capabilities.
- This is equivalent to the RHR mode of operation A-28
- 3. Install a new pump having the higher head and flow
- capabilities while leaving the existing HPSI pumps as is to perform their normal function for the LOCA conditions.
The first possibility was eliminated primarily because the required upgraded flow would be somewhat in excess of the runout flow of the existing pump in combination with the fact that a pump manufacturer advised that directly feeding with a booster pump in this manner would be quite difficult. Further, the savings in cost and ease of installation over a new pump to provide the total head requirement would not be significant.
The third was selected over the second method because of three important considerations:
- 1. Use of separate pumps to meet the TLOFW and LOCA requirements provides a higher assurance of safety.
- 2. It minimizes disruption of the existing HPSI system since it would retain its original configuration and characteristics.
- 3. The costs and time of installing a new pump as compared with replacing an existing pump would not be significantly different.
The last factor to be considered in establishing the basic method for F&B is that of the type of valves to be
.used for the bleed and their functional requirements. The concept of using power-operated relief valves (PORVs) was introduced at the onset of this study. Also, PORVs were considered in the C-E transient analysis (Ref. 1) . Typically, a PORV is either fully open or closed with a fairly fast operating time no provide the relief function. Use of this type of valve wo21d be appropriate for Case 1 where the require-ment is to full, open the valves and leave them open to effect a rapid pressure reduction. Use of a valve of this type in the full F&B system of Case 2 after such time as a1 solid-water system prevails would cause undesirable pressure transients and the pressure would be very difficult to control. Therefore, the use of a modulating type of control valve is considered necessary. The use of a quick-opening relief valve is not considered a requirement for this service. For example, Ref-erence 1 shows the system pressure at the end of the steam gen-erator boil-out interval as approaching the SRV set point of
. 2500 psia at about 2 psi /sec. Thus, for the pressure to increase from a valve opening set point of 2400 psia to a-SRV lift pressure of 2500 psia would require over 45 seconds, which would not require' quick opening of the valves. Final selection of the control valve and actuator types for the bleed se'rvice will require careful investigation because The of the choice largeapressure between quick-difference across the valve. ,
A-29 L -
acting relief valve and a modulating type of control valve -
would depend mainly on the method of injection. If further analysis shows that an existing HPSI Pump would be adequate,
- then.the quick-acting relief valve would be the choice. If
- I a new, full-pressure pump is found to be a requirement, then the choice would be the modulating type of control valve.
3.2 Method Selected for Conceptual Design From the discussions above in Section 3.1, the method selected embodies the following characteristics:
- 1. Starc of the new pump for F&B service at 40 minutes after reactor trip if a TLOFW condition exists and it has been ascertained at that time that AFW cannot be restored. The pump is to inject at a system pressure of 2400 psia with an initial flow rats. of approximately 970 gpm to match decay heat removal at that time by boil-off only. It is to be noted that this selection was made to have a limiting case in terms of pump size but, that in doing a detailed design for a specific plant, transient analyses would be performed to finalize the injection rate.
- 2. Valves controlled to open automatically at 2400 psia. For F&B service this means valves will open at 10 minutes after reactor trip and hold the pressure at 2400 psia until the feed starts at 40 minutes. Valves will be of the modulating control type with special attention given to the large pressure difference across the valve. If an existing HPSI pump can be used, the valves will be of the quick-acting PORV type.
- 3. Discharge of steam, two phase mixture, or water to a new quench tank identical to the existing quench tank. In an extended.TLOFW condition, the tank rupture disk will rupture and permit discharge of fluid to the containment sump. As in the case of the existing HPSI pumps, the new pump will draw its suction. water from the refueling water storage tank (RWST) until this tank is empty and then switch over' to a recirculation mode with water from the containment sump. ,
- 4. After going in*.o operation, the controls for the F&B system will maintain pressure at 2400 psia -
until such. time as the decay heat generation has decreased to the level where the LPSI pumps can take over, which is approximately 3h hours after reactor _ trip. At this time the pump will be shut A-30 t-L
.1 H
i l
. off, the valves fully opened and the pressure allowed to quickly decrease to 150 psia where the LPSI pumps will operate at their design flow rate in the SDC modo.
- 5. Upgrading of the existing IIPSI pumping capability 4
will.bo by means of a new pump with no change in
- the existing IIPSI pumps for LOCA service except the appropriato cross connections for suction and discharge of the now pump.
f l
t J
9.
f l'
i-
- .A-31' d
TABLE 1 FEED & BLEED CASES CONSIDERED
-CASE NO. DESCRIPTION INITIATION TIME-Seconds CORE UNCOVERING BLEED VALVES PUMP
-1 (1) Feed & bleed with 2 new PORVs (5) and 1 existing 1200 2750 (6) Yes HPSI pump 2A - Delayed initiation.with'l new pump sized 3000 (2) 3000 (2) Yes for water heatup 2B - Early Initiation with 1 new pump sized to 600 (3) 600 (3) No
~ prevent core uncovering y ..
.6~ 2CL Delayed initiation with 1 new pump sized 2400 (4) 2400'(4) No w to prevent core uncovering
- NOTES:
- 1. . See Reference 1.
- 2. - Latest . time for restoring auxiliary feedwater flow
- 3. Steam generator dryout time'
- 4. Latest time to prevent core uncovering
- 5. . Large~PORVs - 4" body size 6'. . Time at which actual -injection starts A
O. O O l
r
TABLE 1A FEED & BLEED CASES CONSIDERED (ALTERNATE)
CASE NO. DESCRIPTION INITIATION TIME - Seconds CORE UNCOVERING BLEED VALVES PUMP
~
1 (1) Feed &. bleed with 2.new PORVs (5) and 1 existing 1200 2750 (6) Yes
.HPSI pump 2A-A. Delayed initiation with 1 new pump sized 600 (3) 3000 (2) Yes for_ water heatup -
2B-A - Early initiation. with 1 new pump sized 600 (3) 600 (3) No to prevent core uncovering .
- l' ~ 2C-A ~ Delayed initiation with 1 new pump sized 600 (3) 2400 (4) No tj to prevent core uncovering NOTES:
- 1. See Reference 1
- 2. Latest: time for restoring auxiliary feedwater flow
- 3. . Steam generator dryout time
- 4. . Latest: time to prevent core uncovering
~
- 5. Large PORVs - 4" body size
- 6. Time at which actual injection starts
TABLE 2 FEED & BLEED FLOW REQUIREMENTS CASE NO.- ~ BLEED FLOW - VALVES (total for 2). FEED FLOW - PUMP
. (1). Steam 11bm/sec. Water - gpm Maximam Equilibrium lbm/sec. gpm
~1 240- 4605 712 76 547 2A 127- 2441 9696 78 557 28 '204 4452 1190 95- 684 p.,
2C- 136- 2613 1705 135- 972-T NOTES:
- 1. ' See Table 1 for ' definition of cases 2.: Data from C-E transient; analysis (Ref.1)
I l ... ...
- = . ,
i TABLE 2A FEED & BLEED REQUIREMENTS (ALTERNATE)
CASE NO. BLEED FLOW - VALVES (total for 2) FEED FLOW - PUMP
.(1) Steam - lbm/Sec. .
Water - gpm Maximum Equilibrium 1bm/sec. apm 1(2) 240 4605 712 76 547 2A ' A . 204 4452 969 78 557 28.- A 204 4452 1190 95 684 y 2C - A 204 4452 1705 135 972 j NOTES:
- 1. See Table 1A for definition of cases
TABLE 3 PUMP DATA Estimated Estimated CASE NO. FEED DES'CJi VALUES (2) Nominal No. of Overall Dimensions Brake HP
__ Design Flow-Drive Motor Connected _ Delivered lost- I(4)
(1) Design Head- Casing Design Pump Size Stages Length Width Requi red HP Rating Load - kW Pump Motor gpa _To tal f t of water Press-psia in. in. in. (3) 1 547 2500 1950 - - - - -
600 4 80 - - -
2A 557 5730 3000 6 x 8 x 10% 13 205 54 1092 1250 1000 330.000 68.560 398.560 28 684 5820 3000 6 x 8 x 10h 13 205 54 1313 1500 1200 330.000 80.050 410.050 2C 972 6120 3000 6 x 8 x 12 11 235 70 1787 2000 1600 370.000 102.950 472.950 D'
s E.a cn NOTES:
- 1. See Table 1 for definition of cases
- 2. Refer to Table 2
- 3. Based on motor efficiency of 94%
4 Delivered costs, not including installation
I TABLE 4 C_0ST AND CONNECTED LOAD COMPARISON FOR NEW PUMP FOR FEED AND BLEED SERVICE i
- CASE NO. TOTAL DELIVERED COST INCREASE CONNECTED INCREASE IN LOAD OVER 1
COST OF PUMP-$ OVER BASE CASE. LOAD-kW BASE CASE 5 % kW %
2A 398,560 - - 1000 - -
l (base case) 2B 410,050 11,490 3 1200 200 20
- 2C 472,950 74,390 19 1600 600 60 i
4 A
i 4-e
., 6 t
A TABLE 5 BLEED VALVE DATA CASE NO. BLEED REQUIRED PER VALVE (2) VALVE AVAILABLE FOR (1) Steam Water BODY SIZE BODY SIZE SHOWN 5 I"-
lbm/hrx10 C s 9Pm-max C y C C s s 1 4.32 180 2303 37 4 230 152 2A 2.29 95 1221 19 4 230 152 28 3.67 153 2226 35 4 230 152 2C 2.04 85 1307 21 4 230 152 Y
u NOTES:
- 1. See Table 1 for definition of cases.
- 2. Refer to Table 2
r.
TABLE 5A BLEED VALVE DATA (ALTERNATE CASES)
' CASE NO. BLEED REQUIRED PER VALVE (2) VALVE AVAILABLE FOR BODY SIZE (1) Steam Water BODY- SIZE BODY SIZE SHOWN 5
lbm/hrx10 C s 9Pm-max. C y in. C Cy s
I '4.32 180 2303 37- .4- 230 152 2A 3.67' 153 2226 35 4 230 152 2B '3.67 ~153 2226 35 4 230 152-2C: :3.67 153 2226 35 _4 230 152 Y
e-
, NOTES:
1 See Table 1A for definition of cases
- 2. iRefer to Table 2A L
4.0 CONCEPTUAL DESIGN 4.1 Design criteria In addition to the functional requirements described ,
in Section 3.2, the F&B system will conform to the following design criteria:
- 1) Classifications for AMSE-Code Class, Nuclear Safety Class, Quality Group Class and Seismic Category will be as shown in Table 6.
- 2) The system will have the ability to remove decay heat and hold primary system pressure constant, thus avoiding lifting of the SRVs, while in the hot standby condition.
- 3) The system will have the capability to cool down and depressurize from the hot standby condition to a pressure and temperature where the SDC mode 3 of the LPSI pumps can be used to complete the cooldown to cold shutdown and to maintain the cold shutdown condition. The injection water source will initially be the RWSTs and, if necessary after an extended period of operation, recircula-tion from the.. containment emergency sump.
- 4) Existing HPSI pumps will not be used in the cool-down from hot standby to cold shutdown conditions.
- 5) The system will be capable of operating when off-site and unit turbine-generator power sources are not available provided that power is available from one emergency diesel generator.
l l 6) Redundant F&B systems or equipment will not be l required.
i 7) System design will permit inspection and testing on a periodic basis under conditions as close to design requirements as possible.
- 8) Instrumentation and control will be provided in the existing control room, where a new separate panel will be used.
- 9) Modification and rearrangements of existing structures, -
equipment and piping will be minimized.
- 10) System will be designed with the objective of minimum '
plant outage time to install, preferrably to be done within normal plant outages for refueling or sched-led maintenance.
A-40
TABLE 6 CLASSIFICATIONS FOR EQUIPMENT AND PIPING ASME NUCLEAR QUALITY- SEISMIC ITEM CODE-CLASS- SAFETY CLASS GROUP CLASS CATEGORY FEED SIDE
- 1. NEW PUMP III-2 2 B I
- 2. ISOLATION VALVES, INSTRUMENTS, III-2 2 B I CONTROL & OTHER APPURTENANCES
- 3. PUMP SUCTION PIPING III-2 2 B I
'4. PUMP DISCHARGE PIPING UP TO III-2 2 B I-CHECK VALVES
.l . '5. CHECK VALVES CONNECTING PUMP III-2 2 B I
"" -DISCHARGE TO RCS
- 6. PIPING' FROM CHECK VALVES TO HEADER III-2 2 B I BLEED SIDE
'l. BLEED VALVES III-I I A I
- 2. ISOLATION VALVES FOR BLEED VALVES III-l 1 A I
- 3. QUENCH TANK III-3 3 C I
- 4. INSTRUMENTS, CONTROL AND OTHER III-3 3 C I
' APPURTENANCES
'5. PIPING FROM PRZ TO BLEED VALVES III-l 1 A I
- 6. PIPING FROM BLEED VALVES TO QUENCH TANK III-3 3- C I III-3 3 C I
- 7. QUENCH TANK DRAIN PIPING
- __= - ._
- 11) The system shall be designed to prevent exposure of '
equipment and piping to pressures and temperatures which exceed design limits. ,
- 12) The system will be designed not to interfere with or jeopardize other safety systems during normal or abnormal conditions.
- 13) In the event of a LOCA signal while the F&B systent is operating, the F&B system will be secured and the normal SIS functions will take over.
- 14) Initiation of F&B system operation will be automatic to the extent that unanticipated automatic actuation cannot cause or exacerbate accident conditions.
- 15) Manual control will be possible to override automatic system control or to reactivate the other safety systems.
- 16) The system will be designed to prevent simultaneous operation of the new pump with the existing HPSI and LPSI pumps.
- 17) Pressurizer heaters will not operate during F&B system operation.
- 18) Charging pumps will not operate while P&B system is in operation.
- 19) Water injected by the F&B system will not require additional boration beyond the level of boration in the TWST.
- 20) Instrumentation will be provided to sense leakage past the system bleed valves. .
4.2 System See Figure 1 for the F&B system and Table 7 for its com-ponents.
The F&B system is a single-train system with its equip-ment powered from Diesel Generator B.
Suction for the pump is taken from the existing 24 inch lines from the RWSTs T-005 and T-006, which supply the existing HPSI and LPSI pumps, through motor-operated isolation valves V6, -
V11 & V12. Each of the RWSTs has a capacity of 245,000 gallons.
Since the P&B system is intended to operate for a period of 3 hours until the SDC mode of the LPSI pumps takes over, and A-42
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l i TABLE 7 f FEED & BLEED SYSTEM' COMPONENTS l ITEM NO. COMPONENTS .
l P1 High pressure safety injection pump Tl Quench tank T2 Pressure damping tank V1 Bleed valve No. 1 V2 Bleed valve No. 2 V3 Motor-operated isolation valve - Bleed Valve No. 1 V4 Motor-operated isolation valve - Bleed Valve No. 2 V5 Deleted V6 Motor-operated isolation valve - new pump suction i
V7 Motor operated isolation valve - new pump discharge to RC loop 1A V8 Motor operated isolation valve - new pump discharge to RC loop 1B
! V9 Motor operated isolation valve - new pump discharge to RC loop,2B V10 Motor operated. isolation valve - new pump l
l l
discharge to RC loop.2A l
l Vll Motor operated isolation valve - new pump_
l
, suction from RWST.1 l
V12. Motor operated isolation _ valve - new' pump -
l
- suction from RWST 2 ,
i
'A - _
TABLE 7 (Continued)
ITEM NO. COMPONENTS vl3 Quench tank N2 pressure control valve V14 Quench tank makeup water valve V15 Quench tank drain valve V16 New pump discharge header pressure relief valve V17 Pressure damping tank N 2 Pressure control valve V18 Manual isolation valve - new pump discharge V19 Check valve - new pump discharge V20 Manual isolation valve - quench tank N 2 pressure line V21 Manual isolation valve - Pressure damping tank N
2 Pressure line V22 Isolation valve - breakdown orifice 2 V23 Isolation valve - breakdown-orifice 3 V24 Check valve - breakdown orifice 1 V25 Check valve - breakdown orifice 2 V26 Check valve - breakdown orifice 3 01 Breakdown orifice 1 02 Breakdown orifice:2 03 Breakdown orifice 3 A-45
.. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _a
the maximum flow rate of the new pump is less than 1000 gpm, ,
the contents of one RWST would suffice. The LPSI pumps would then have their suction valves aligned to take water from the containment emergency sump. However, if required, the F&B .
system could take water from the second RWST.
The new pump discharges through four motor-operated isolation valves, V7-V10, to feed into the existing 8-inch lines from the LPSI pumps discharge to RC loops lA, 1B, 2A &
2B. Feeding of all four RC loops is consistent with the existing HPSI and LPSI system configurations.
During a F&B operation the fluid discharged through the bleed valves will start as 100% quality steam and then progress through an increasing liquid component of two-phase mixture until a solid water situation prevails. Under this condition, system pressure will be very sensitive to changes in the position of the valves; therefore, tank T2 is connected at the pump dis-charge to dampen pressure transients resulting from the valve control action. Nitrogen pressure in this tank is controlled by a self-acting pressure control valve, V17.
The combination of motor-operated valves, V22 & V23, check valves, V24, V25 & V26, and pressure breakdown orifices, 01, 02
& 03, are provided for control of recirculation flow from the pump discharge to the RWSTs. These will be used to limit the pressure in the pump discharge header. Control will be by means of' pressure measured at the pump discharge header. As a backup to avoid excessive pressure in the discharge piping, a safety relief valve, V16, is located on the pump discharge header. This recirculation arrangement will also permit on-line testing of the pump without actual injection.
The bleed function is accomplished by two identical valves, V1 & V2, connected at the pressurizer top as are the existing SRVs. Normally-open motor operated valves, V3 & V4, can be used to isolate either bleed valve should excessive leakage develop through the valves. Temperature and acoustic sensors in the discharge lines from the valves will warn of. leakage. The dis-charge lines run to a quench tank T1, with characteristics -
similar to those of the existing quench tank connected to the discharge of the SRVs.
Self-acting pressure control valve V12 serves to maintain a constant nitrogen cover gas pressure in the quench tank.
Control valve V14, is used to admit makeup water to the -
tank from the reactor makeup pumps. Control valve V15 is used to control level in the quench tank by draining to the '
reactor drain pumps on high level in the tank.
A-46
. The quench tank, with a capacity of 2300 gallons, is not intended to contain the discharge during an extended TLOFW situation but rather to contain any blowdown from pressure transients. or valve leakage. In an extended TLOFW condition the tank rupture disk would open at 125 psig under the increasing steam pressure from the blowdown. After a period of releasing steam throug the ruptured disk, the quench tank would fill with water and overflow with the discharge ultimately going to the containment sump.
4.3 Components Major components in the F&B system are the new pump, the two bleed valves and the quench tank. These are described in the sections which follow.
4.3.1 New Pump ,
The pump characteristics are summarized in Table 8 and an outline drawing is shown in Figure 2.
l The new pump would be a horizontal, centrifugal barrel-type, multistage pump driven by direct coupled electric motor. Pump and motor drive mounted would be located in the existing No. 2 Safety Injection Pump Room at elevation - 15'6".
The pump would take suction from either existing Refueling Water Tank T005 or T006 through ponnections to suction headers feeding the existing Safety Injection System. Each suction line to the new pump would be provided with motor operated isolation valves operable from the Control Room. The pump discharge would be provided with check valve and motor-operated isolation valve and routed from the No. 2 Safety Injection Pump Room to l the existing Low Pressure Safety Injection Header in the piping penetration area outside of containment. Supporting steel and l pipe hangers would be provided for suction and discharge piping runs. Motor-operated isolation valves would be provided at each of the four connections to existing low pressure injection lines outside of containment penetrations. A safety relief valve would be provided to protect existing piping and equip-ment from over-pressure.
All equipment, piping, foundations and supports would be Nuclear Class-2, Seismic Category 1.
- Complete indication of pump pressure, valve conditions and discharge pressure would be provided in the Control Room.
4.3.2. Bleed Valves The bleed valve cnaracteristics are summarized in Table 9 and an outline drawing is shown in Figure 3.
A-47
FIGURE 2 1
PUMP OUTLINE i .
l l
l WH - : A = c K MINIMUM FOR DISMAN TLING
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l DIMENSIONS IN INCHES NO A B C D E F G H J K STAGES 7
13 121'4e 65h 30%4 25'/s 18 36 54 99 2044 86'/s 1 .
l A-48 l
l
TABLE 8 NEW PUMP
SUMMARY
SPECIFICATIONS Identification. P1 Number 1 Type 1-Stage, Horizontal Centrif-ugal, 2500 lb Discharge Flange, 150 lb Suction Flange, Austen-etic Stainless Steel, Forged Casing, Mechanical Seals Nominal Size 6 x 8 x 10h In.
Rated Capacity 972 GPM Rated TDH 6100 Ft.
NPSH 20 Ft.
Speed 3600 RPM Drive Direct-Coupled Electric Motor Bearings Thrust & Journal, Oil Cooled Coolant Component Water ,
0 Des. Press & Temp, 2750 PSIG, 200 7 Special Requirements See Table 6 Motor Details Type Squirrel Cage Induction (may require reduced voltage starting)
Enclosure Drip - Proof i Rated-H 2,000 Cooling Air'
-Special Requirements See Table 6
+
A-49
TABLE 9 .
BLEED VALVES ,
Identification- V1 & V2 Number 2 Capacity 150 lbs/sec Sat. Steam @
2400 gSIA; 9500 GPM Water
@ 662 F Type .
Case 1 - Solenoid-actuated PORV Case 2 - Drag. valve with air cylinder.or operator Size 4" Service Controlled Throttling Body 316 Stainless Steel Ends Weld Plyg 316 Stainless Steel -
(Hardened Seat & Plug)
Design Press. 2670 PSIA Design Temperature 700 F ASME Class 1 Seismic Category 1 Special Features-Provision-for stem position
' indication d
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.A-50~
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A-51 ,
l Two bleed valves would be installed with new piping, each one being connected to one of the existing 6" inlet lines -
to existing pressurizer safety valves. Six inch motor-operated isolation valves would be installed between each valve and safety
- valve inlets functioning as bleed valve shut-offs.
For Case 1 the bleed valves would be of the PORV type in which an electrical solenoid receiving a signal from an external pressure measurement channel actuates a pilot-oper-ated valve for the main relief valve. This type of valve is curre tly in use on some PWRs in a similar application.
For the full F&B system of Case 2 the bleed valves would be straight through flow, air-actuated pressure control valves under automatic control or control room operator's control. Figure 3 shows a drag type of control valve with air cylinder operator which is in use in certain PWR units for pressurizer bleed valve service. The drag valve type is especially suitable for controlling under high pressure drop conditions. Each would have a valve position indicator affording positive means of valve position indication in the control room.
Capacity of each would be 150 lbr.'sec of steam 100% satgrated at 2400 psia and 9500 GPM of water at 2400 psia and 662 F saturated temperature. They would discharge through a common discharge header and submerged suppression piping inside the F&B quench tank.
Valves and piping supports as well as access stair-way and platforms would be provided for maintenance and testing of the bleed valves. All equipment, supports and structures would be Seismic Category 1. Piping would be Nuclear Class 1 from pressurizer through the bleed valves and Class 3 beyond.
4.3.3 F&B Quench Tank The tank characteristics are shown in Table 10.
A new F&B quench tank having similar design char-acterisitics to the existing pressurizer quench tank would be provided. Final design would be established by detailed engineering. The tank would be located on added supporting structures inside the confines of the existing biological shield at approximate elevation 61'5". The tank would be designed to receive and quench bleed valve discharge for transient pressure excursions or valve leakage. Relieving capacity would be provided by a top-mounted tank rupture disk. .
Pressure controlled N 2 Supply would be provided by extension of the existing N 2 supply system via a self-actuated .
pressure control valve.
A-52
- Also, by extension of existing services, control of the quench tank makeup and high level drain would be provided.
Level control, indication and alarms would be extended to the Control Room.
The quench tank would be provided with access manhole, access platform and structural supports designed for Nuclear Class 3 and Seismic Category 1.
Bleed eguipment and materials would be prefabricated to the maximum practicable extent and installed and tested during a normal plant shutdown period.
TABLE 10 F&B QUENCil TANK Identification Tl Number 1 Capacity 2300 Gal.
Installation Vertical lieight & Diameter 10' x 8' Shell Material 316 Stainless Steel Shell Thickness 3/8", Heads 1/2" Corrosion Allowance ASME III Code Stamp Reg'd. ASME III, Class-3 Special Features Internal Perforated Suppressor - distributor Nuclear Class 3 Seismic Category 1
- For estimating purposes only.
Design to be finalized by. detailed
- Engineering.
A-53 4
4.4 Arrangement of Equipment and Piping -
The general arrangement for the new pump on the feed side of the F&B system is shown in Figure 4. This pump
- would be located in the room with the existing No. 2 HPSI pump at floor elevation (-) 15'-6" in the auxiliary building.
A dismantling length allowance has been made for the new pump.
and access for installation will be through the overhead access hatch at floor elevation 8' -
0". Pump accessories for the new pump, the pressure damping tank, recirculation flow control valves, motor operated suction valve, and manual isolation
~
valve and check valve on the discharge side will all be located in the No. 2 pump room.
All suction and discharge piping for the new pump will be installed outside of containment; thus, there will be no need for new containment piping penetrations. A single 6" discharge line will be run from the new pump to a header from which connections will be made through motor-operated isolation valves to the four existing 8" lines from the LPSI pumps to reactor coolant loops lA, 1B, 2A & 2C. The LPSI lines were chosen over the HPSI lines to the reactor cooling loops because of the size advantage (8" versus 2").
Connections to the existing lines from the new pump discharge will be made between the existing isolation valves and the containment penetrations. This is necessary because the new pump discharge requires a piping design pressure class con-sistent with the high-pressure reactor coolar.c piping. There is a pressure class change at the isolation valves from the LPSI low-pressure level to the high-pressure level of the RCS. A flow nozzle to measure net flow of the new pump to the reactor will be installed in an appropriate straight-run section of the pump 6" discharge line. In its intended normal mode of operation the new pump would be required to operate from the point of reactor trip after a TLOFW until the SDC mode of the LPSI pumps can take over the decay heat removal function, a period of about 3 hours. This would require the contents of only one RWST; however, to give the P&B system the flexibility for extended operation beyond 3 hours, connections will be made from both RWSTs to the new pump suction. This will require making a connection to each of the existing 24" lines through a motor-operated isolation valve, and thus through a single 6" line to the pump suction.
The new pump will have the same assurance of NPSH as the existing HPSI pumps.
Arrangement of the new quench rank and valves on the bleed side of the F&B system is shown in Figure 5. .
A-54 m_
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A-57 f
All bleed side equipment and piping will be located within the biological shield walls around the pressurizer. .
Two new valves and their associated motor-operated isolation valvec will be installed at the top of the pressurizer from the diametrically opposite pipe stubs on the pressurizer top -
head which connect to the existing SRVs. Lines from the PORVs will join to form a common 12" line down to the new quench tank.
The existing pressurizer quench tank is horizontally mounted above the elevation 45' - 0" floor level within the biological shield walls making up the pressurizer enclosure.
The new quench tank will have the same size and shape as the existing tank but be mounted vertically above the elevation 63' - 6" floor in the pressurizer enclosure and above the existing tank. Since the elevation 63' - 6" floor does not presently extend inside of the pressurizer cubicle, it will be necessary to install steel framing to support the new tank and an access floor.
Overflow through the rupture disk of either the new or existing quench tanks will flow to the pressurizer cubicle floor and eventually to the containment sump.
4.5 Electrical 4.5.1 Feed and Bleed System Electrical Loads The electrical loads which may be required during operation of the new F&B System are listed on Table 11. Although the total load is 4680 kW it should be noted that it includes two reserved loads which may not become actual loads during the course of the feed and bleed operation. The first reserved load of 656 kW is for the auxiliary feedwater pump. The inclu-sion of this load gives an assurance that the auxiliary feedwater pump can be started immediately, if the cause of subminimum feedwater flow is eliminated. The second reserved load of 332 kW is for the containment spray pump. The pump would be put into operation manually by the operator if the pressure inside con-tainment reaches the allowable limit. The pump operation may not be required at all since the containment emergency and upper dome circulating fans should be able to keep the contain-ment pressure under the limit.
Since the new pump start-up will result in one step load increase of 1600 KW, it will have to be determined that the resulting inrush current is still within the diesel generator's
- capability. Some auxiliary equipment for bringing the new pump into initial motion may have to be considered. Also, ,
reduced-voltage starting may be a requirement.
A-58
TABLE 11 ELECTRICAL LOADS REQUIRED FOR FEED & BLEED OPERATION $
5
. VOLTAGE LOAD 5 DESCRIPTION FUNCTION (VOLTS) (kW) M New Pump Provides Water for Core Cool- 4160 ~1600 12 ing Circulate Seawater From 4160 331 Salt Water Cooling Pump Intake Structure to the CCW Heat Exchanger Provide Cooling Water To 4160 478 4 Component Cooling Pump NSSS Equipment Aux. Feedwater Pump Provides Feedwater to the 4160 656 6 Steam Generators 4160 482 7 Aux. Bldg, Emer. Chiller Provides Chilled Water for Emergency A.C.
Provides Cooling Spray in 4160 332 M Containment Spray Pump Containment Control of the New Feed and 480 50 11 New Motor Operated.
Valves / Instrumentation Bleed System and Controls Provides Containment Stru- 480 121 1 Containment Emergency Fan ture Air Cooling and Recirculation Provides Cooling to the 480 47 2 Upper Dome Air Circu-lating Fan Containment by Circulating Dome Air Pro'vides Cooling to'the 480 98 2 Control Room Emer. AC Unit Control Room O
A TABLE'll (Cont'd.)
u
- 5 VOLTAGE LOAD 8 '
DESCRIPTION FUNCTION .
(VOLTS) (kW) W Miscellaneous HVAC Provides HVAC for FHB, Con- 480 100 2 trol Room, SWGR Room, Battery Room, Etc.
Diesel Generctor Provides HVAC and Other 480 165 1,3,4 Support Systems Services for Diesel Gener-ator Motor Operated Valves Opens and Closel Valves as 480 40 1 Required Battery Chargers Provides DC Power and 480 150 1 Maintains Battery at Charged State -
Instrumentation / Provides Power for Inst. & 480 30 1 Controls / Monitors Control / Measures Radiation Levels -
Total 4680 Sequence Starting Time (Excludes 10 seconds allowable time for for the diesel generator to come up to speed and voltage after the start signal. Zero time ~is when the diesel generator circuit breaker closes.)
(Seconds) 1 0 2 5-3 10 4 15 5 20 - -
6 30 7 35 '
11 600 or more 12 600 or more M Manual-A-60
E e
4.5.2 Electrical System Design The one-line diagram for the complete F&B System electrical equipment is shown in Figure 7 (notes and symbols are in Figure 6).
To accommodate the new F&B System electrical loads it will be necessary to install a new unit substation with a 4160 - volt supply from the existing vital bus B. This is the same bus that supplies power to existing HPSI and LPSI pumps No. 2, which will be in the same cubicle as the new pump, and receives emergency power from diesel-generator B. The unit sub-station of metal clad switchgear will contain a 4160 - volt circuit breaker and disconnect switch for the feeder from the existing vital bus B, a 4160 - volt circuit breaker (s) for the new pump, a 4160/480 - volt transformer, a motor control center, and an instrument compartment.
The 480 - volt motor control center will include motor starters for the following:
. o~ All new motor-operated isolation valves o New pump auxiliaries (if required) o Supplementary air-conditioning unit for the pump cubicle (if required).
Power for the new F&B instrumentation and control will also be supplied from the 480-volt bus.
Starting of the new pump is an item of special concern since it will be much larger than any of the existing loads supplied by the diesel generator (e . g . 2,000 Hp vs.
600 Hp for existing HPSI pump). Although the total loads to be connected to the diesel-generator during F&B operation will be within the generator nameplate rating, the motor inrush current may be excessive with respect to the unit's transient capabilities. This would be further aggravated by the fact that the large newpump would be loaded on the diesel bus after a significant portion of the normally sequenced loads.
To mitigate this starting current problem, it may be necessary to use reduced voltage starting. For a motor of this size, Wye-Delta (closed transition type) starting would probably be a requirement. This would limit inrush current to 33% of that with full voltage starting. The reduced ;
starting torque and consequent longer acceleration time would l not be considered objectionable in this application. For re-duced-voltage starting the unit substation would include.the additional circuit breakers, transitional and control equipment required.
9 9
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A-61
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l Por the case of new PORVs but with an existing HPSI pump, the unit would be as for the complete F&B system with 4160 - volt circuit breaker (s), control and instruments for the new pump omitted. .
The dedicated panel for P&B in the main control would include ammeters, alarms, control switches and status lights for the electrical equipment.
The balance of the electrical work would include the running of exposed conduits (and possibly some tray work addition) cables and wiring to accommodate the substation feeder, the 4160 - volt new pump motor and the various 480-volt loads.
In order to accommodate the new pump in the diesel -
generator loading, some modification of the existing controls and wiring for load sequencing would be required.
4.6 Instrumentation & Control 4.6.1 F&B Control Logic for the F&B system control is shown in Figure 9 (notes and symbols are in Figure 8).
The Feed.& Bleed System Actuation Signal (F& BAS) is generated by the following two conditions:
- 1. The primary system pressure is at or above normal and
- 2. The feedwater flow is less than minimum.
The F& BAS will set off the new audio / visual alarm in the control room, open the new valves, start the new pump a'nd energize other equipment. There can be prede-termined time delays (e.g., 2400 seconds) employed subsequent to the F& BAS and prior to the opening of the new valves, the startup of the new pump and energizing of other equipment.
There will be manual override allowing the elimination of any or all time delays.
After the initial few minutes of pump operation, the new depressurization system will operate in the feed and bleed mode maintaining the primary _ pressure between 2100 and 2200 psia in order to. keep the core in~subcooled condition. 'Three-and a. half hours after shutdown the existing Shutdown Cooling (SDC) mode of the LPSI pumps is capable'of core heat removal
- at an acceptable level and the;new pump can be' turned off _
manually. The new valves would be manually closed about ten to-twenty seconds later when the primary pressure drops to about- ,
170 psia and the-exist'ing SDC mode would take over.
A-64
- - _ _ _ - _ _ _ _ _ _ _ _ _ _ - - _ - _ _ _ A
FIGURE 8 CONTROL LOGIC NOTES & SYMBOLS 1r LOCAL SENSOR LOGIC "NOT"(INVERTER) l (SEE NOTE 1) (SEE NOTE 5) 1r I
- luut BLSTABLE '
AND TRIP POINT (SEE NOTE 1) () (SEE NOTE 5)
If it it ir 1r jur_3r COINCIDENCE LOGIC LOGIC "N AND" 2/3 (SEE NOTE 2) (SEE NOTE 5) 1r 1r TIME DELAY UU LOGIC "OR" TD OR (SEE NOTE 3) (SEE NOTE 5) 1r II 1r IP S R LATCH LOGIC CONTINUATION TO OR FROM ANOTHER O (SEE NOTE 4)
V PAGE.
I A-65 l
."t FIGURE 8 NOTES ,,
- 1. Analog voltage signals from process measurement sensors are transmitted to bistable voltage comparators. The bistable compares the input signal voltage to a predetermined setpoint voltage level. When the setpoint is reached, a trip signal is produced (output is a logic 1).
- 2. Coincidence logic generates a coincidence output signal upon the receipt of the minimum required bistable trip signal inputs.
The minimum inputs required are 2 out of 4 or 2 out of 3 desig-nated by 2/4 and 2/3 respectively. Thus when 2 or more bistable trip signal inputs are received, a coincidence output signal is produced (output.is a logic 1).
- 3. The time delay momentarily delays the coincidence output signal to the input of the initiation latch by a predetermined period of time.
- 4. The purpose of the latch is-to maintain the actuation output signal upon receipt of an initiation signal from the coincid-ence logic. This actuation output signal (output is a logic 1) is maintained until the latch is reset. Operation of the latch is defined by the following truth table:-
. Set S t
0 1 0 1 t =-present time t-1 =-time immediately Reset R 0 0- 1 1 t prior.to present-Output O t-1 1 0 1
-5. Logic " NOTE" output.is'the inverse ~of input.
"AND" all inputs must be-present to produce an output.
"NAND" 'all inputs must be present:to' remove output.
"OR" any one input produces an output.
A-66.
l n e
FIGURE 9 CONTROL LOGIC DIAGRAM BLEED VALVE CONTROL PRIM ARY FEEDWATER I PftESSURE FLOW
< MINIMUM S P3 3 PI (2100 PSI) (2200 PSI)
AND AND 4
NY Y 0
MANUAL MANUAL N RESET RESET
! R S R S R i
j INITIATION LATCH INITIATION LATCH O O l MANUAL MANUAL ACTU ATION ACTUATION
- \
DECREASE BLEED VALVE INCRE ASE BLEED VALVE FLOW FLOW A !
FIGURE 9 CONTROL LOGIC DIAGRAM F & B INITI ATION LOGIC PRIMARY FEEDWATER PRESSURE I FLOW I -
1r if if IP
) Pi < NORM AL AW Up
$ P2 > NORMAL (2200 PSI) . #
(200 PSI)
FLOW FLOW 1 r1F MANUAL RESET AND If g r 1f if S R INITI ATION LATCH O
1 r IF 1f AUDIO / VISUAL ALARM 1r 1r if TD. TD TO MANUAL MANUAL MANUAL ACTUATION ACTUATION ACTUATION U J L1' OR OR 03 BLEED BLEED PUMP PUMP VALVEI VALVE 2 START ISOL. VALVES OPEN OPEN A-68 OPEN
If the minimum feedwater flow is restored after the new F&B system is put into operation, the feed and bleed
, will be aborted and steam generators used for decay heat removal.
. In the very incredible case of loss of all feedwater and subsequent LOCA the new F&B system would operate only if the LOCA is small enough, so that the primary system pressure can be maintained at 2100 psia. Should the pressure drop below that, the new P&B system will be turned off and the existing ECCS will take over when the pressure drops below 1600 psia.
An automatic control system has been described above. In actually carrying o detailed design, it may be desirable to consider the alternative of manual control, especially for Case 1 where new PORVs would be used with an existing HPSI pump. Possible reasons for considering manual control are:
o Relative simplicity of the control and decresed potential for spurious initiation o Elimination of the need to interface with existing primary pressure and feedwater flow instrumentation, which would prevent jeopardizing existing safety--
related instrumentation o Provision of operator choice to maintain full primary pressure to the latest possible time after TLOFW for possible restoration of auxiliary feedwater.
4.6.2 Instrumentation Reference is made to Figure 1 for the instrumentation to be included in the F&B System. Summarized below are the points of measurement and their application;
- a. All motor-operated isolation valves-position:
o Status indicating lights - control room j l
o Digital status monitoring - computer l
- b. Bleed Valves - position:
! o Position indicators - control room o Digital status monitoring - computer o "Open" alarm in control room if automatic or manual controls require closed position A-69
- c. Pump-net injection flow:
o Recorder - control room o Analog monitoring - computer
- d. Pump - recirculation flow o Recorder - control room o Analog monitoring - computer
- e. Pump-suction pressure:
o Indicator - control room & local' o Low alarm - control room
- f. Pump-discharge temperature o Indicator - control room o High alarm - control room
- g. Pump-header discharge pressure:
o Recirculation valves control o Recorder - control room o Indicator - local o High & low alarm - control room o Analog monitoring - computer
- h. Pump motor - status:
-o On-off-trip lights - control room o Digital status monitoring - computer
- 1. Pump motor - bearing & winding temperature o High alarms - control room o Analog monitoring - computer
- j. Bleed valve discharge lines -~ temperature and acoustic monitoring.for stuck open valves o Indicator . control room A-70 L . .. .. . . . .
_ _ __ _ _ a
o High alarms - control room
- k. Quench tank - level o Makeup & drain valves control o Indicator - control room o High alarm - control room o Analog monitoring - computer
- 1. Quench tank - pressure o Indicator - control room & local o High & low alarm - control room
- m. Pressure damping tank- pressure o Indicator - control room & local o High & low alarm - control room Primary system pressure and loss of feedwater signals
-as required for F&B system control will be obtained from existing plant instrumentation channels.
A dedicated control panel will be located in the control room to include all indicators, indicating lights, recorders, controllers, control switches and alarms associated with the F&B system.
Analog and digital signals from the F&B system will be connected as inputs to the-existing computer system.
4.7 Structural Structural work required to install the F&B system consists of.a base for the new pump, supports,-flooring and l
walkways for the new quenchLtank, and pipe supports for the new piping. Structural details used for estimating purposes l
are shown in-Figures 10 and ll for the feed and bleed sides I respectively.
l Referring to Figures 5 & 11, steel framing and-a rein-forced concrete floor are called for at elevation 63' - 6" for-support of.the new quench tank. Additionally, walkways and
- i. platforms are required'to provide access to the rupture disk
' at the' top of the. quench tank.
L L
A-71
FIGURE 10 STRUCTURAL DETAIL- NEW PUMP FOUNDATION FEED SIDE 210 7/is" g . . ..
NEW PUMP BASE PLATE g 54" WIDE 8 204 7/id' LONG l l i
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I A-72
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FIGURE 11 STRUCTURAL DETAIL- PL ATFORM AT ELEV. 63'-6" BL8ED SIDE' e
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A-74
5.0 IMPLEMENTATION
- 5.1 Installation Installation of a full F&B system in an operating plant would be planned and carried out with the following objectives:
o Performance of as much as possible of the work prior to the plant annual shutdown in order to minimize the balance of work required during the shutdown period o No interference with normal plant operations o No reduction of the function, availability or integrity of existing plant safety or engineered safeguards systems o No unnecessary radiation exposure to installation personnel.
Because of limited time available during plant shut-down, the difficulties imposed by spatial arrangement constraints in an existing plant, and the possiblity of personnel radiation exposure, the use of a scale model or mockup for critical areas is. considered essential for the purposes of installation planning and training. Such areas include:
! o The valves, quench tank and piping in the area of the pressurizer o The cubicle housing the existing HPSI and LPSI pumps and to house the new pump o The area where the discharge line from the new pump will connect to existing injection lines to the primary coolant loops'through isolation valves o The area where the suction line to the new pump will connect to existing safety injection pump suction lines through isolation valves.
Structural preparation would be relatively simple since no major structural modifications would be required. For the feed side, the preparation would include providing a rein .
forced concrete base for the new pump, installation of pipe support brackets and piping penetration through the pump cubicle wall. On the bleed' side, structural preparation would be somewhat more extensive. It would require structural steel framing and concrete floor to support the new quench-tank, A-75
)
walkways and platforms to provide access to the top of the quench tank, and pipe support brackets. Furthermore, all of .
this would have to be done during the early part of the shut-down prior to installation of the quench tank and connecting piping. -
All major feed and bleed piping would be shop fabri-cated and pre-assembled, using spool pieces where required, to ensure dimensional compliance and ease of installation.
Prior to plant shutdown the new pump would be placed and the suction and discharge piping installed, insofar ac possible, to the point of tie-in with existing piping. It is expected that partial disassembly of the new pump (e.g., base-plate, pump and motor) will be a requirement because of pump unit and access hatch size. Further disassembly, such as pump casings and rotating parts, motor stator and rotor, etc. may be necessary but should be avoided if at all possible.
Installation critical path items will be primarily the bleed side equipment in the area of the pressurizer and secondarily, the tie-in of the new pump suction and discharge to existing piping.
The installation of new electrical switchgear, conduit, trays, cable and wiring will be mainly outside of containment and can be mostly done prior to plant shutdown'throughout the entire construction schedule. Only power supply tie-in to essential switchgear and interconnections with the existing diesel load sequencing control may have to be deferred to the shutdown period.
The same is true for instrumentation and control with respect to the control panel and cubicles, most of the local instruments and controls, conduit, wire and tubing.
Instrumentation and control installation work within contain-ment for the bleed side, which is a minor portion of the total work,-as well as tie-ins with existing' instrumentation channels and circuits, may have to be deferred to the shutdown' period.
5.2 Testing-Testing for the F&B sy3 tem would have to be conmen-surate with that normally provided for plant engineered safeguards systems. These. tests would be in the categories of preoperational tests and system operational test procedures.
Preoperational tests would include the following:
- o. System. hydrostatic testing o Electrical circuit tests '-- continuity and insulation resistance A-76 i
o Circuit breaker and motor starter operation o Instruments and controls - operation, control and calibration o Alarm circuit operation o Motor rotation o Valve operator actuation, including limit settings System operational tests, which would have to be performed in the critical second half of the scheduled outage would include:
o New pump operational flow tests using the pump recirculation line o Operational tests of the emergency diesel generator loading sequences as modified for the F&B system o F&B initiation logic control in response to simulated TLOFW and low primary pressure signals In addition to operational tests on the F&B. system itself, operational tests would have to be performed on existing plant systems and equipment to assure that the functioning and integrity of these systems had not been adversely affected where interfaces were made with the new system. These would include interfaces with the following:
o Primary system pressure instrumentation channels o Diesel-generator load sequencing control o Plant computer system o Interlocks to assure that the new pump and and existing HPSI or LPSI pumps could not be powered siimultaneously.-
During plant operation periodic tests would be per-formed to check the response to simulated TLOFW and high pressure signals to_ check the: system logic through to opening of the bleed valves with their associated isolation _ valves closed. Also, _ periodic tests would :be performed to check the
. new pump in a recirculation mode with suction taken from the RWSTs. This would be done with? the isolation valves from pump discharge to the reactor: coolant loops closed to prevent actual l flow injection into the primary system.
l A-77 l
l <
During the engineering and design phase of the work, detailed procedures for pre-operational, startup and periodic on-line testing will be prepared by plant test and
- operations personnel working closely with the cognizant design personnel.
Simulator programming and training for the plant operators will be modified to incorporate the operational char-acteristics of the F&B system and its testing.
5.3 Costs Costs for the implementation of F&B capability were estimated for four different situations:
- 1. Installation'in a new plant during the latter stages of construction of two new PORVs to provide the bleed capability and use of an existing HPSI pump for feed
- 2. Installation in a plant that has been in oper-ation for some time of two new PORVs and use of an existing HPSI pump
- 3. Installation in a new plant during the latter stages of construction of two-new bleed valves and a new pump
- 4. Installation in a plant that has been in oper-ation for some time of two new bleed valves and a new pump.
A summary of the costs for each of the above cases is shown in Table 12. These were prepared by separately estimating construction costs and the costs for supporting services on a present-day basis. Construction costs include mechanical, structural, electrical and instrumentation and control work.
Cost for services are mainly those for engineering, design and analysis; however, they also include project management, planning, scheduling and cost estimating, quality assurance, construction management, testing and startup and training.
Present-day cost estimates were escalated at an annual rate of 6%, assuming that the project would not start for one year and that the average of the total cash flow would occur at 18 months into the project (see Section 5.4). Interest during construction was based on a 12% annual rate for two years at an average amount of 50% of the escalated costs. An ,
-overall contingency allowance of 25% was made because of a number of uncertainties in plant - specific situations such as radiation -
levels in working areas.
A-78
For the cases of a new plant in the latter stages of construction, it has been assumed that the decision to implement a P&B system would be made at a time when the plant design had been essentially finished and construction had pro-
. ceeded to the point where the system would have to be considered a retrofitting operation. Thus, essentially the same design and construction work would be required whether in a new plant or an operating plant. The cost differentials would arise from con-sideration of difficulties and time required (e.g., as a result of radiation exposure to personnel) in the plant that had been in operation for some time.
Cost estimates include allowance for the higher labor costs incurred by the burnout of personnel working in high radiation areas (see Section 5.5).
The estimates of Table 12 show that the total costs for a complete F&B system can be expected to run 2.5 to 3 times the costs for installing new PORVs only with use of an existing HPSI pump. Installation of new PORVs only in an operating plant are estimated to run about 35% higher than in a new plant; and for a complete F&B system, about 17% higher. The lower increase in the latter case reflects the fact that a larger fraction of of the work could be done outside of containment.
Two cost factors in a plant-specific situation that could very significantly increase the total implementation costs that are not included in the ettimates of this study are:
- 1. Replacement energy costs for a total plant outage time for system installation in an oper- i ating plant that would exceed the normal annual I outage time for refueling and scheduled mainten-ance ,
1
- 2. The need for a new, dedicated diesel-generator l to supply power for the system.
The first of these two factors would affect implemen-tation in operating plants only. Furthermore, it would depend to a large extent on the actual annual outage duration and how well the work was planned and executed to complete it within this time. For purposes of this study an annual outage duration of two months was used. It is considered possible to accomplish the required installation work during this period by very careful planning and doing as much installation as possible prior to actual shutdown. Increased outage time to do the work, either as a result of a normal annual outage of less than two months or unforeseen installation difficulties in installation, could double the P&B inplementation costs in a week or two of extended outage with replacement energy costs charged.
A-79
TABLE 12
. COST ESTIMATE
SUMMARY
-SHEET 1 -'NEW PORVs WITH EXISTING HPSI PUMP INSTALLATION IN NEW PLANT UNDER CONSTRUCTION .
ITEM ESTIMATE COST - $
- 1. Construction 1.1 Mechancial Equipment & Piping 665,000 1.2 Structural 35,000 1.3 Electrical 27,000 1.4 Instrumentation & Control 236,000 Total Construction 963,000
- 2. Services 2.1 Project Management, Planning & 52,000 Scheduling.& Cost Estimating
~
2.2 Engineering, Design & Analysis 423,000 2.3 Quality Assurance 20,000 2.4 Construction Management 40,000 2.5 Test and_Startup' 20,000 2.6 Training 18,000-Total Services 573,000
- 3. Total Present Estimated Costs 1,536,000
- 4. Escalation 246,000 Sub-Total l',782',000
- 5. Interest.During. Construction 214,000 Sub-Total- 1,996,000-
- 6. Contingency 499,000 7 . Total Estimated Costs at Completion. 2,495,000 A-80 T
- b. '
__m'... - 5
! TABLE 12 l
COST ESTIMATE
SUMMARY
q SHEET 2 - NEW PORVs WITH EXISTING HPSI PUMP INSTALLATION IN OPERATING PLANT ITEM ESTIMATED COST - $
- 1. Construction 1.1 Mechanical Equipment & Piping 1,132,000 1.2 Structural 126,000 1.3 Electrical 117,000 1.4 Instrumentation & Control 556,000 Total-Construction 1,931,000
- 2. Services 2.1 Project Management, Planning & 65,000 Scheduling & Cost Estimating 2.2 Engineering Design and Analysis 424,000 2.3 Quality-Assurance 24,000 2.4 Construction Management 48,000 2.5 Testing & Startup 24,000 2.6 Training 58,000 2.7 Health Physics 45~,000 Total Services 688,000
- 3. Total Present Estimated Costs 2,619,000
- 4. Escalation 419,000 Sub-Total 3,038,000
- 5. Interest'During Construction 365,000 Sub-Total 3,403,000
- 6. Contingency 851,000
- 7. Total Estimated Costs at Completion 4,254,000
'9 A-81
TABLE 12 COST ESTIMATE
SUMMARY
SHEET 3 - NEW BLEED VALVES WITH NEW PUMP INSTALLATION IN NEW PLANT UNDER CONSTRUCTION -
ITEM ESTIMATED COST - S
- 1. Construction 1.1 Mechanical Equipment & Piping 2,463,000
'. 1.2 Structural 77,000 1.3 Electrical 202,000 1.4 Instrumentation & Control 424,000 Total Construction 3,166,000
- 2. Services 2.1 Project Management, Planning & 98,000 Scheduling and Cost Estimating 2.2 Engineering, Design & Analysis 851,000 2.3 Quality Assurance 40,000 2.4 Construction Management 64,000 2.5 Testing & Startup 28,000 2.6 Training 36,000 Total Services 1,117,000
- 3. Total Present Estimated Costs 483,000
- 4. Escalation 686,000 Sub-Total 4,969',000 5.- -Interest During Construction 597,000 Sub-Total 5,566,000
- 6. Contingency 1,392,000
- 7. Total Estimated Costs at Completion 6,958,000 A-82 l
TABLE 12 COST ESTIMATE
SUMMARY
- SilEET 4 - NEW BLEED VALVES WITH NEW PUMP INSTALLATION IN OPERATING PLANT ITEM ESTIMATED COST - $
- 1. Construction 1.1 Mechanical Equipment & Piping 4,054,000 Structural 173,000 1.2 Electrical 259,000 1.3 Instrumentation & Control 574,000 1.4 Total Construction 5,060,000
- 2. Services Project Management, Planning & 124,000 2.1 Scheduling and Cost Estimating 2.2 Engineering, Design &' Analysis 852,000 2.3 Quality Assurance 48,000 2.4 Construction Management 72,000 2.5 Testing & Startup 32,000 2.6 Training 75,000 2.7 !!calth - Physics 85,000 Total Services 1,288,000 Total Present Estimated Costs 6,348,000 3.
1,016,000
- 4. Escalation Sub - Total 7,364,000 Interest During Construction 884,000 5.
Sub - Total 8,248,000 2,062,000
- 6. Contingency
- 7. Total Estimated. Costs at completion 10,310,000 ih ', A-83 i
Addition of the bleed valves would require that primary system hydrotesting be performed. In a situation where refueling is in the critical path this could add two or three days to the outage, meaning that replacement energy .
costs for this time could be chargeable to the addition of the system. On the other hand, the turbine-generator is frequently on the critical path during the annual refueling outage. .
For SONGS the ability of an existing emergency diesel-generator to supply the required F&B loads was found to be a marginal situation. This results mainly from the fact that a new pump will require more than twice the electrical power of an existing HPSI pump. In other plant-specific situations there is a distinct possiblity that an existing diesel-gener-ator would not be adequate and the installation of a new, dedicated diesel-generator would be required. A possible mit-gating factor is the indication in the Argonne analyses (Ref. 9) that, in the case of TLOFW and the loss of off-site power, an existing HPSI pump with new PORVs may be sufficient.
Although the estimating of costs for such a contingency was not a requirement of this study, earlier work done by Burns and Roe for Sandia on an Alternate Decay Heat Removal (ADHR)
(Ref. 7) system study would indicate an approximate cost for diesel-generator, auxiliary systems and equipment, and structure in the range of $10 million to $12 million. This could very well predominate all other costs for F&B implementation.
5.4 Schedule An overall schedule for the implementation of P&B capability, a schedule for engineering, design and analysis, and a construction schedule are shown in Figure 12. This schedule is based on the most stringent case of installing a complete F&B system, consisting of new bleed valves and pump, in a plant that had been in operation for some time.
The other cases would have less demanding schedules.
A time span of 18 months is allowed for engineering, design and analysis, and 12 months for construction, with an overlap of 6 months between the two, resulting in an overall schedule of 2 years. Two factors would govern in establishing the overall schedule end pointc:
- 1. The engineering, design and analysis phase could not begin until a sufficient base of rulemaking and criteria had been promulgated by the USNRC to permit the detailed design work to proceed with a reasonable assurance of compliance. Typical of items that would have to be included in such a base are:
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r CONTROLLED DEPRESSURIZATION FOR PWRs j FIbURE 12 SHEET 1 - IMPLEMENTATION SCHEDULE OVERALL ACTIVITY
- 1. NRC RULEMAKING I
- 2. ENGINEERING, DESIGN 8
& ANALYSIS I Y
$ 3. CONSTRUCTION i
3 I
Scheduled Annual b Plant outage t e
- 4. FSAR ADDENDUM # '
a f I 1 0 6 12 1B 2'4
____A -
PfGU8F f2 SHEET F . IMPtfMf 4 TAT!0N SCHf 00tf - EN".f 4fERING. Of5f GN & ANAlv515 ACTIVITY
- 1. ACQUI$! TION OF PLANT DESIGN INFO.
2, P9EFARATION OF 0[51GN CRITERIA i -
- 3. ANALYSE 5 o FULL LO M REJECTION REQUIREMENT 5 H A
o THERM 0HYORAut !C/TRAN51ENT o ELECTRICAL LOA 0!NG
- BLEED VALVE TYPE 5 ELECTION ; '
a a o RA0!ATION 00$E s e o PIPING ( STRESS. PIPE WHIP ETC.) ,
o 5E!1MIC , ,
5 y 4 MECHANICAL DESIGN o SYSTEM > LOW O!AGRAM l l o EQUIPMENT ARRANGEMENT h 8
o PIPING ARR/NGEMENT o PIP!NG 150 METRICS l INSTALLATIO't DETAILS } !
o . ,
o VALVE & EQUIPMENT LISTS
- 5. 57RUCTURAL DESIGN
. i o STRUCTURAL DETAILS
- 6. ELECTRICAL DEstGN a
o ONE-LINE O!AGRAM g o INSTALLATION DETAIL 5 .,
a a o CACLE/WIlt!NO LISTS o GROUNDING PLAN
- 7. INSTR. & CONTR.
=
o INSTR. $CHEMATICS E y
o CONTROL LOGIC o CONTROL PANEL LAYOUT o th5TRUMENT L15T ,
o INSTALLATION DETAIL 5
- 8. SPECIFICATIONS. 810 EVALVATION
& ORDER PLACEMENT o HEW PUMP 0 '
- 8LEED WALVE5 ' '
o VALVE 5. TANK 5 & MISC. ,
5 MECHANICAL o PIP!NG ,
INSTR & CONTROL 5 i o {
o ELECTRICAL SWITCHGEAR g
o MECHANICAL INSTALLATION o ELECTR. & INSTR. INSTALLATION g
- 9. $AFETY ANALY515 REPORT .
s y g
, 3
- 10. SCHEDULING B I
- 11. COST ESTIMATING I i 4
i I I C 6' 12 is ELAPSED MONTHS A-86
i f
FIGURE 12 SHEET 3 - IMPLEMENTATION SCHEDULE - CONSTRUCTION ACTIVITY
- 1. STRUCTURAL PREPARATION l h
- 2. PIPING FABRICATION 1 1
- 3. NEW PUMP PLACEMENT & PIPING f INSTALLATION
- 4. PUMP SUCTION AND DISCHARGE l CONNECTIONS TO EXSISTING PIPING
- 5. MECH. INSTALLATION OF VALVES ,
QUENCH TANK & PIPING
- 6. ELECTRICAL INSTALLATION
- 7. INSTR. & CONTROL i INSTALLATION '
- 8. SYSTEM TESTING q Scheduled Annual
lant outage k
I 1 _
12 18 24 i A-87 ELAPSED MONTHS l
o Requirement with respect to the single failure criteria (e.g., need for redundancy
~
o Definition of exactly what accident conditions .
apply such as TLOFW, small LOCA and steam gener-ator isolation .
o Functional relationship between existing ECCS and F&D system o Requirement for operation under loss of off-site and on-site normal power supply.
For purposes of estimating cost escalation only a one-year period for formulation of this base has been allowed. This is purely nominal and in no way represents an estimate.
- 2. Completion of installation would be planned to coin-cide with the end of the second annual plant outage following the start of design work. In the event of inability to complete installation during the annual outage, there may be an option to complete the work during the next annual' outage rather than incurring the penalty of replacement energy costs.
The schedule of engineering, design and analysis in-cludes the general activities of plant design information acqui-sition, preparation of design criteria, a number of specialized analyses, detailed design in the various disciplines, and_the preparation of specificiations, bid evaluation and equipment order placement.
Based on advice from a typical pump manufacturer, a period of 13-14 months would have to be allowed between placement of order and delivery. Thus, the pump would be on the critical-path and any analysis required to establish its size would have to be assigned.an.up front priority. The other critical procure-ment item, although not possibly as critical as that of the pump, would be the PORVs or bleed valves. These valves, which would be in Nuclear Safety Class 1 and be located at.the reactor
. coolant pressure boundary, may require time for qualification testing superimposed on the normal procurement time. Here again, any special analysis to select-the-valve. type and size would have an up-front priority.
The construction schedule shown is based on carefully.
planning and doing as much of the work as possible prior to' plant-shutdown, subject to radiation exposure limits to. personnel and-non-interference with normal plant operations and engineered safe- ;
guards systems availability.- For example, the feed side of the .
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system which lies entirely outside of containment could be worked by placing the pump and fabricating and installing.the suction and discharge piping up to the final points of connection with
- existing piping. Also, much of the electrical and instrumentation work'outside of containment could be accomplished with the plant in operation.
5.5 Radiation Exposure
. One of the major considerations in installing-upgraded
- F&B capability in an-operating plant will be that of radiation exposure to personnel. Two principal areas to be occupied by personnel during the installation work that are of concern are
- 1. The area around the pressurizer where the new valves, quench tank and associated piping would be installed
- 2. Outside of containment, the cubicle containing existing HPSI and LPSI pumps where the new pump would be installed ( No. 2 cubicle in SONGS).
Reference 8, which shows average radiation levels at shutdown in a number of PWR plants which have been in oper-ation for a number of years, contains the following information:
o 0.4 R/hr maximum in the area of the pressurizer spray line o 0.3 R/hr maximum contact level at the RHR pumps (functional equivalent of the SONGS LPSI pumps in the SDC mode of operation).
These levels in a plant-specific situation will depend on a number of factors such as:
o Length of time the plant has been in operation
, o History.of failed fuel o Corrosion product buildup in the primary coolant o Amount of usagelof RHR or SDC o Primary coolant surges-in~and.'out of pressurizer.
, Shutdown ~ radiation levels in the pressurizer'and the pumps used for RHR cn SDC will depend' on the. deposition or. l plate-out of long-lived ~ fission and corrosion products.
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Work on the bleed side of the F&B system, consisting of installation of the valves, quench tank and associated piping inside primary containment in the pressurizer area, is -
assumed to be essentially all done while the plant is shutdown because of the high radiation level from short-lived radio-nuclides (e.g., N-16) during plant operation. To estima'.e *:
the number of burnouts expected during work in this area, the above-cited radiation level of 0.4 R/hr is used. This is c'on- ;
sidered conservative since it is a maximum at the pressurizer f spray line and would be less in other parts of the pressurizer i i
area; also, because credit is not taken for use of temporary shielding since that may be difficult in particular situations.
As a basis for estimating the number of burnouts ,
exposure of a worker to the maximum 0.4 R/hr level during an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shift would use up the quarterly allowance of 3 Rem. For each person thus using up his quarterly allowance, an allowance J of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> training for a replacement is made plus allowance i' for qualified personnel to provide the training.
On the feed side of the F&B system, the primary concern would be for exposure resulting from radiation levels at the "
LPSI pump adjacent to where the new pump is to be installed. The above-cited maximum of 0.3 R/hr and the location and configu-ration of the LPSI pump is considered such that the use of ,
temporary shielding would provide cufficient attenuation to preclude burnout concerns. 'l A very important aspect of planning this work is to 0 hhve a health physicist involved in the project right from the 3 start. Initially he would be responsible for acquiring radi- 3 ation maps and other data for both the oeprating and shutdown Then he would be involved in the installation plant conditions.
planning to minimize and control radiation exposure; and finally to monitor this during the actual installation.
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6.0 OPERATION AND MAINTENANCE 6.1 Operation Principal Modes of Operation
. 6.1.1 Startup Prior to plant startup, the Feed and Bleed (P&B)
System is isolated from the reactor coolant system, and aligned for actuation with instrumentation calibrated and operating.
The F&B system set points are defeated until criticality is achieved.
6.1.2 Normal Operation During normal operation, the F&B system is in a standby mode. The set points are maintained at their design points.
6.1.3 Shutdown During shutdown the F&B system is in a standby mode.
The set points are maintained at their design points.
6.1.4 Special or Infrequent Operation 6.1.4.1 Alternate Water Sources Although the normal source of water (i.ei two RWSTs -
245,000 gallons each) is adequate for the designed function of the P&B system, provisions are made to switch the new pump suction ta) the containment sump. Thus the F&B system con-tinuous operation well beyond the nominal'31s hours is assured.
6.1.5 Emergency 6.1.5.1 Loss of Off-Site Power Upon loss of off-site powar, the diesel generator automatically starts up to supply - Tical power for the safety related equipment. If an F&b s /dtem Actuation Signal (F& BAS) is generated, the non-essential auxiliaries are shed to provide adequate startup margin for the new HPSI pump motor.
6.1.5.2 Feed and Bleed System Actuation Signal (F& BAS)
See section 4.6.1.
6.2 Maintenance Provisions will be made for flushing the borated A-91
water from the P&B system to prevent corrosion caused by stagnant borated water in the stainless stcol components.
The new PORV or biced valve or its internals, depending on the valve design selected, will be replaced during the refueling outage following the accumulated use
- of the P&B system of more than 3h hours.
The new pump and motor will have a mainter.ance schedule identical to the existing ilPSI pumps.
6.3 Testing The new pump will be periodically tested during plant operation in a manner similar to that used for the existing liPSI and LPSI pumps by recirculating the pump dis-charge to the RWSTs.
Each PORV or bleed valve will be tested periodically for opening and closing during plant operation with its associ-ated motor-operated isolation valve closed. Actual blowdown testing during plant operation will not be used. Valves will be removed, inspected, bench-tested and, if necessary, refurbished during the annual refueling outage.
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7.0 REQUIREMENTS FOR DETAILED ENGINEERING,
, DESIGN AND ANALYSIS The conceptual design of a F&B system developed during this study would serve as the basis for proceeding with the a detailed engineering, design and analysis necessary for installing such a system in a given plant. This study has also served to identify a number of special investigations and analyses that would have to be performed in conjunction with the normal detailed engineering and design activities.
Areas where special investigations and analyses would be required include the following:
- 1) Further transient analyses of the type performed by C-E (Ref.1) would be necessary to optimize the design in terms of the time after reactor trip that P&B should be inititated, what maximum injection flow rate would be required to prevent core uncovering, and what the resulting pump horsepower and electrical load requirements would be. Of particular impor-tance here would be accurate or demonstrably con-servative modeling of core thermohydraulics to define the extent of the two phase fluid mixture zone and its effect on fuel clad temperatures. The approach used in carrying out the conceptual design for SONGS was to use a latest time of initiation consistent with what was evaluated to be a " worst case" injection flow rate, based on the existing transient analysis, in order to conservatively assess the technical feasibility and costs. A more specific and carefully modeled transient analysis of an upgraded F&B system could lead to less stringent pump flow and power requirements.
- 2) Analysis of the electrical loadang of the emergency diesel-generator for an upgraded F&B system in SONGS indicated a very marginal situation as far as steady-state load versus diesel-generator rated kW capacity is concerned. Such a situation for a specific PWR to be upgraded would require a very careful analysis of what loads are essential during a F&B operation.
Also, the effect on the diesel-ganerator of a large locked-rotor inrush current at the instant of starting the large new pump would have to be investigated.
- 3) Planning for the installation of that portion of the F&B system to be located within primary contain-
- ment (e.g. , in the area of the pressurizer) will require measurement of the radiation levels that will exist after shutdown insofar as it will effect the A-93 i
a permissable exposure time of personnel doing the installation.
- 4) In addition to the normal stress analysis required ,
for the design of piping and supports, special analysis will be required to determine the effects of thrust loads associated with the high blowdown .
rates and the effects of a pipe break in the new blowdown piping.
- 5) Because of the high pressure drops and changes in fluid conditions (steam to two-phase to solid water) involved, considerable investigation of the type of valve to use for the PORVs or bleed valves will be required. This may require some supporting thermo-hydraulic anal,ysis. /
- 6) Detailed planning will be needed to accomplish installation of the system coordinated with the annual outage period scheduled for refueling and maintenance.
Normal documentation required for the detailed engineering and design'of the system would include:
o Piping and Instrumentation Diagram o Electrical One-Line Diagram o Control Logic Diagrams o Instrumentation Schematic Diagrams o Piping Arrangement Drawings o Piping Isometric Drawings o Equipment Arrangement Drawings o Structural Details o Mechanical Installation Details o Control Panel Layout o Valve, Instrument and Equipment Lists o Electrical Installation Details o Procurement Specifications for Pump, Valves,-
Tanks, Electrical Equipment, Instrumentation ,
and Controls
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o Safety Analysis Report Addendum 5
. o Project Schedules o Detailed Cost Estimates i
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8.0 APPLICABILITY TO OTHER PLANTS This study was performed with the objective of devel- .
oping a system for controlled depressurization that would be generally applicable to PWRs lacking this capability.
However, to investigate the feasibility of installing a system .
and the associated costs, SONGS was selected as a typical nuclear power plant.
The system conceptual design described in Section 4.0 would be, in principle, applicable to any PWR requiring a means for controlled depressurization. In designing and installing a system for a specific plant, certain factors that depend on variations in plant design parameters and features would have to be taken into consideration. These are discussed below.
The first of these considerations would be concerned with the size of the PORVs or bleed valves required and at what time after a TLOFW they should be opened; and whether an existing HPSI pump would be adequate to prevent core damage, and, if not what injection flow rate and size of new pump would be required, as well as when injection flow should start.
Determination of these feed and bleed system basic requirements would depend on such plant parameters as reactor power level rating, primary coolant volumes and secondary cooling system volumes. The amount of decay heat to be removed by a feed and bleed system will vary directly with reactor power level rating. Primary coolant system volumes and geometry of flow paths will effect the time for injected cool water to reach the core. Secondary cooling system water volumes, particularly in the steam generators, would affect the. dry-out time for the steam generators.
Transient analyses of the type that have been performed by CE and ANL would be required to determine the basic require-ments for the feed and bleed system for a specific plant taking into account its parameters as described above. How-ever, before doing this it would be necessary to have a firm set of criteria for protection against fuel damage as applicable to all plants. For example, such criteria could be based on the conservative approach of no core uncovering as determined by the collapsed water level; or by the less conservative ;
level defintion oI where the two-phase mixture becomes 100%
steam. !
The feasibility and cost of installing a system for control-led depressurization in a specific plant will depend to a large -
extent on whether or not an existing emergency diesel generator can meet the electrical power requirements of the system. This , l would only apply if a new, larger pump were found to be necessary.
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Since the diesel generator rating and loading for SONGS may be fairly typical of other plants, combined with the fact that
- use of a conservatively large new HPSI pump was still within the diesel-generator capacity, there is a reasonable probability that c' er plants of similar design would not require a new,
, dedicated diesel-generator.
Another-important consideration is the availability of space to locate a new pump in the event that such a cumn is required. This should be as close to containment'as pos-sible, preferably in one of the existing safety-injection pump compartments. The situation in SONGS where one of the pump com-partments has sufficient room to accommodate another pump may not be typical of other plants. In a plant where the relocation of equipment and/or piping was found necessary to accommodate a new pump the implementation cost could be significantly higher than estimated for SONGS.
For the case where a new pump is required, the size, pressure rating and location of existing safety injection piping will be another important factor in applicability to a specific plant. In the case of SONGS, a favorable combin-ation of sufficiently large piping and access for connections outside of containment in piping sections having a high pressure rating were found. This is not necessarily a typical situaticn that could be expected in all plants under consideration.
Arrangement of equipment and piping arcund the pressurizer are a major concern to be considered in a specific plant. Avail-ability of a suitable pressurizer connection for the bleed-valves is essential. If space for a dedicated quench tank for the PROVs is found to be a problem, there is the possibility of utilizing the existing quench tank which receives blowdown' from the SRVs.
As far as the type of bleed valves.to be used is:cencerned, once this has been established-for the first plant, it should be
. generally applicable for other plants.
Adding a system which requires PORVs.cnly, as compared with a. system wnich requires a new pump as well'as bleed valves, will generally be more readily applicable to other plants because of the less. stringent requirements for equipment space, piping accommodations'and electrical power supply.
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REFERENCES
- 1. Combustion Engineering, " Decay Heat Removal and Depressurization, Performance Evaluation, Total Loss of Feedwater Transient Analysis", January 1983 .
(Analysis results only - not a formal report)
. 2. San Onofre 2& 3 FSAR
- 3. Palo Verde FSAR (Units 1, 2 & 3)
- 4. American Nuclear Society, ANSI (ANS) 5.1-1979,
" Decay Heat Power in Light Water Reactors"
- 5. USNRC, Regulatory Guide 1.139, " Guidance for Residual Heat Removal", May 1978
- 6. Letter of March 3, 1983 from D. Erickson of Sandia to E. Gahan of Burns and Roe.
- 7. Burns _and Roe, Inc., " Alternate Decay Heat Removal Concept Evaluation", February 1982.
- 8. Battelle Pacific Northwest Labs. , NUREG/CR-0130, Volt. 2,
" Technology, Safety and Costs of Decommissioning a Reference
. Pressurized Water Reactor Power Station," June 1978.
- 9. Argonne National Laboratory, "TLOFW Transient Analysis",
(Analysis results only - not a formal report)
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APPENDIX B MAJOR DETAILS OF THE RELIABILITY ASSESSMENT DONE FOR SONGS 2 e
B-1 1
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TABLE OF CONTENTS Page 1.0 Introduction ............................................ B-3 1.1 AFWS Recovery ........................................... B-3 ,
l.2 SONGS Dominant Accident Sequences without PORVs ......... B-4 1.3 SONGS Dominant Accident Sequences with PORVs ............ B-8 l
3 i 1.4 SONGS Dominant Accident Sequences with a Feed and l Bleed System ............................................ B-10 ;
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f B-2 i .. . - - . . .- .. .-- . - - . . - . . .- -. . . -
1.0 INTRODUCTION
This appendix provides additional details of the analysis done to determine the benefits which would be obtained by adding a feed
, and bleed capability to SONGS 2. However, it is not intended to provide sufficient detail to allow reconstruction of the entire analysis. The first issue addressed is recovery of the AFWS at SONGS given that some failure occurred. Then, the dominant accident sequences, i.e., sequences with a frequency greater than 1.0 E-7 per year, for the three cases of feed and bleed capability investigated for SONGS are discussed in some detail.
1.1 AFWS RECOVERY The estimated unreliabilities for the SONGS AFWS with and without recovery are presented in Table 1.1. When loss of offsite power (LOP) is the initiating event and the AFWS fails on demand, there is an 80% chance it will be recovered within 20 minutes and an 85% chance it will be recovered in 50 minutes. When loss of the MFWS is the initiating event and the AFWS fails on demand, there is a 94% chance it will be recovered within 20 minutes and a 96% chance it will be recovered within 50 minutes. Considered another way, when LOP is the initiating event and the AFWS fails initially and is not recovered in the first 20 minutes, there is then a 27% chance it will be recovered within the next 30 minutes. Likewise, when loss of the MFWS is the initiating event and the AFWS fails initially and is not recovered in the first 20 minutes, there is a 33% probability it will be recovered in the next 30 minutes.
Table 1.1: Unreliability of the SONGS 2 AFWS AFWS AFWS AFWS Reliability Reliability Reliability Initiating Without With Recovery Event With Recovery Recovery (20 min.) (50 min.)
Loss of Offsite 5.4 E-4 1.1 E-4 8.0 E-5 Power Loss of MFWS 5.6 E-5 3.3 E-6 2.2 E-6 The significance of the above probabilities can be seen when considering operator actions to mitigate a transient by using PORVs and the HPSI system for feed and bleed (see Figure 4.3, sequence 17). In this case, the operator must make a decision to feed and bleed at 20 minutes although 50 minutes are presumed available for recovery of the AFWS. Under these conditions the operator might be reluctant to initiate feed and bleed for various reasons with the hope that the AFWS would be recovered in the remaining 30 minutes. However, the above reliabilities indicate B-3
that there is not a good change that the AFWS could be recovered in these 30 minutes, and consequently the operator should follow procedures and go to a feed and bleed mode of operation to prevent core melt.
1.2 SONGS Dominant Accident Sequences Without PORVs There are six accident sequences at SONGS 2 which have an .
estimated core melt frequency greater than 1.0E-7: T IML, T2ML, S2EAD, S2RMED, (SGTR) RAD, and (SGTR)RMED. These sequences are discussed below.
1.2.1 Sequence T l ML The sequence IT ML is initiated by LOP (T 1) which causes the simultaneous loss of the MFWS (M). This is followed by loss of the AFWS (L) and failure to recover either offsite power or the AFWS within 50 minutes. The frequency of this sequence is estimated to be:
f(TI ML) = 7.0E-6 per year.
The dominant cut sets of this accident sequence are listed below.
Cut Set Frequency
- T1*DGA-GEN-LF*BATB-BAT-LP 5.3E-6 (.25)
T1*DGB-GEN-LF*BATA-BAT-LF 5.3E-6 (.25)
T1*AFWP140X-PTD-LF*DGA-GEN-LF*DGB-GEN-LF 2.6E-6 (.25)
T 1*AFWP140X-P-MNTP*DGA-GEN-LF*DGB-GEN-LF 1.6E-6 (.25)
T 1* BAT-CM 9.0E-7 (.25)
T1*AFW-ACT-A-L*AFW-ACT-B-L 4.4E-6 (.03)
- Multiplying the cut set frequency by the number in parentheses gives the frequency with recovery.
B-4
Term Descriptions Tl: loss of offsite power; f(TI) = .09 per year t . DGA-GEN-LF: local fault in diesel generator A; P(DGA-GEN-LF) = 5.4E-2 DGB-GEN-LF: local fault in diesel generator B; P (DGB-GEN-LF) = 5.4E-2 BATA-BAT-LF: local fault in station battery A; P(BATA-BAT-LF) = 1.lE-3 BATB-BAT-LF: local fault in station battery B; P (BATB-B AT-LF) = 1.lE-3 BAT-CM: common mode failure of the station batteries; P(BAT-CM) = 1.0E-5 AFWP140X-PTD-LF: local fault of the AFWS turbine pump; P ( AFWP14 0X-PTD-LF) = 1.0E-2 AFWP140X-P-MNTP: maintenance on the AFWS turbine pump; P(AFWP140X-P-MNTP) = 6.0E-3 AFW-ACT-A-L: loss of the AFWS A-train actuation; P(AFW-ACT-A-L) = 7.0E-3 AFW-ACT-B-L: loss of the AFWS B-train actuation; P ( AFW- ACT-B-L) = 7.0E-3 1.2.2 Sequence T 2ML The sequence T 2 ML is initiated by loss of the MFWS (T2M).
This is followed by loss of the AFWS and failure to recover the AFWS in 50 minutes. The frequency of this sequence is. estimated to be:
f(T 2ML) = 2.6E-6 per year.
B-5
l The dominant cut sets of this accident sequence are listed below:
Cut Set Frequency .
T 2*AFW-ACT-A-L*AFW-ACT-B-L 5.9E-5 (.03)
Term Descriptions T2: loss of the MFWS; f(T2) = 1.2 per year AFW-ACT-A-L: loss of the AFWS A-train actuation; P ( AFW- ACT- A-L) = 7.0E-3 AFW-ACT-B-L: loss of the AFWS B-train actuation; P(AFW-ACT-B-L) = 7.0E-3 1.2.3 Sequences S RAD 2 and (SGTR) RAD The sequences S2 RRD and (SGTR) RAD are initiated by a small LOCA and steam generator tube rupture, respectively. Core melt follows due to failure of the HPSI system (D). The frequency of these sequences is estimated to be:
f [ (IE) R AD] = 3.8E-6 per year, 1
where IE signifies the initiating event.
The dominant cut sets of these accident sequences are listed below.
Cut Set Frequency IE*R*HPIO68-XOC-LP 1.8E-6 (1)
IE*R*HE-ACT-RESET 1.8E-5 (.05)
IE*A*TM-LOGIC-CM 1.8E-6 (.05) .
B-6
Term Descriptions IE: initiating event; f(S2) = f(SGTR) = .02 per year HPIO68-X0C-LP: plugging of a manual valve in the HPSI minimum flow recirculation line; P(HPIO68-XOC-LF) = 1.0E-4 HE-ACT-RESET: failure to reset the HPSI actuation system after testing; P(HE-ACT-RESET) = 1.0E-3 TM-LOGIC-CM: common mode failure of the HPSI system actuation logic due to test or maintenance; P (TM-LOGIC-CM) = 1.0E-4 N: success of the MFWS; P(R) = 0.9 1.2.4 Sequences S RMED 2 and (SGTR)RMED The sequences S2 RMED and (SGTR)RMED are initiated by a small LOCA and steam generator tube rupture respectively. The initiating event is followed by failure of the MFWS (M), success of the AFWS (L), and failure of the HPSI system (D) . Core melt results from failure to make-up RCS inventory. The frequency of these sequences is estimated to be:
f [ (IE) RMED) = 3.8E-7 per year.
The dominant cut sets of these accident sequences are listed below.
Cut Set Frequency IE*M*HPIO68-X0C-LP 2.0E-7 (1)
, IE*M*HE-ACT-RESET- 2.0E-6. ( .05) l IE *M *TM-LOG IC-CM 2.0E-7 (.05)
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Event Descriptions IE: initiating event; f(S 2 ) = f (SGTR) = .02 per year HPIO68-X.0C-LF: plugging of a manual value in the HPSI minimum flow .
recirculation line; P (HPIO 68 -XOC-LF) = 1.0E-4 [
HE-ACT-RESET: failure to reset the HPSI actuation system after testing; P (HE- ACT- RESET) = 1.0E-3 TM-LOGIC-CM: common mode failure of the HPSI system actuation logic due to test or maintenance P (TM-LOGIC-CM) = 1.0E-4 M failure of the MFWS; P (M) = 0.1 1.3 SONGS Dominant Accident Sequences with PORVs There are five accident sequences for SONGS 2 with PORVs which have an estimated core melt frequency greater than 1.0E-7:
T IRMLPD, S2RRD, S2RMLD, (SGTR) RRD, and (SGTR) RMED. These sequences are discussed below.
1.3.1 Sequence T KMLPD 1
The sequence TKMLPD is initiated by LOP I(T ) which results in MFWS failure (M). This is followed by failure of the AFWS (L) without recovery in 50 minutes and failure of the HPSI system to provide make-up to the RCS for the feed and bleed operation (D).
l The estimated frequency of this accident sequence is:
l l
f(T 1RMLPD) = 5.5E-6.
I the dominant cut sets of this accident sequence are listed l below.
B-8
Cut Set Frequency
- T 1*DGA-GEN-LF*BATB-BAT-LF 5.3E-6 (.25)
. T1*DGB-GEN-LF*BATA-BAT-LF 5.3E-6 f.25) 4 T1*AFWP140X-PTD-LF*DGA-GEN-LF*DGB-GEN-LF 2.6E-6 (.25)
T 1*AFWP140X-P-MNTP*DGA-GEN-LF*DGB-GEN-LF 1.6E-6 (.25)
T1* BAT-CM 9.0E-7 (.25)
Term Descriptions T1: loss of offsite power; f(T 1) = .09 per year DGA-GEN-LF: local fault in diesel generator A; P(DGA-GEN-LF) = 5.4E-2 DGB-GEN-LF: local fault in diesel generator B; P ( DGB-GEN-LF) = 5.4E-2 BATA-BAT-LF: local fault in station battery A;-
P(BATA-BAT-LF) = 1.lE-3 BATB-BAT-LF: local fault in station battery B; P (BATB-BAT-LF) = 1.lE-3 AFWP140X-PTD-LF: local fault in the AFWS turbine pump; P ( AFWP14 0X-PTD-LF) = 1.0E-2 AFWP140X-P-MNTP: maintenance.on the AFWS turbine pump; P(AFWP140X-P-MNTP) = 6.0E-3 1.3.2 Sequences S 2KMD and . (SGTR)KMD
-These sequences are described in Section 1.2.3.
1.3.3 Sequences S9KMID and (SGTR)KMLD I - These sequences are described in Section 1.2.4.
B-9
1.4 SONGS Dominant Accident Sequences With a Feed and Bleed System There is only one accident sequence at SONGS 2 with a feed and bleed system which has a core melt frequency greater than 1.0E-7:
T1RMLF. This accident sequence is discussed below. .
1.4.1 Sequence TIRMLF The sequence T IRMLF is initiated by LOP (TI) which results in MFWS failure (M). This is followed by AFWS failure (L) without recovery in 50 minutes and failure of the feed and bleed system (F). The estimated frequency of this accident sequence is:
f(T RMLF) 1
= 5.9E-6 per year.
The dominant cut sets for this accident sequence are listed below.
Cut Set Frequency T1*DGA-GEN-LF*BATB-BAT-LF 5.3E-6 (.25)
T1*DGB-GEN-LF*BATA-BAT-LF 5.3E-6 (.25)
T 1*AFWP140X-PTD-LF*DGA-GEN-LF*DGB-GEN-LF 2.6E-6 (.25)
T1*AFWP140X-P-MNTP*DGA-GEN-LF*DGB-GEN-LF 1.6E-6 (.25)
T1* BAT-CM 9.0E-7 (.25)
Term Descriptions Tl: loss of offsite power; f(T1) = .09 per year DGA-GEN-LF: local fault in diesel generator A; P(DGA-GEN-LF) = 5.4E-2 DGB-GEN-LP: local fault in diesel generator B; P ( DGB-GEN -LF) = 5.4E-2 BATA-BAT-LP: local fault in station battery A; P(BATA-BAT-LF) = 1.lE-3
- 1 B-10 !
BATB-BAT-LF: local fault in station battery B; P(BATB-BAT-LF) = 1.lE-3
. AFWP140X-PTD-LF: local fault in the AFWS turbine pump; P(AFWP140X-PTD-LF) = 1.0E-2 AFWP140X-P-MNTP: maintenance on the AFWS turbine pump; P ( AFWP14 0X-P-MNTP) = 6.0E-3 i
)
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Distribution:
U.S. NRC Distribution Contractor (CDSI) 7300 Pearl Street Bethesda, MD 20014 l (100 copies for AN)
U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, DC 20555 Attn: T. P. Speis F. Schroeder K. Kniel P. E. Norian A. J. Buslik R. J. Mattson R. W. Houston
. L. S. Rubenstein L. B. Marsh R. Lobel O. D. Parr A. C. Thadani C. Y. Liang l D. G. Eisenhut .
i R. Bollmer H. Thompson s
B. Sheron F. H. Rowsome A. R. Marchese (10)
H. Rood V. Nerses E. Licitra J. Wilson 6400 A. W. Snyder 6410 D. J. McCloskey 6411 A. S. Benjamin 6411 S. W. Hatch j 6411 R. G. Spulak 6412 J. W. Hickman 6412 W. R. Cramond 6412. A. M. Kolaczkowski 6412 G. J. Kolb 6412 A. C. Payne 6412 T. A. Wheeler 6414 S. L. Daniel 6414 D. M. Ericson, Jr.
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DIST-1
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NRC POmu 336 U s NuCLEAm mEcuLATOny commission m en NUREG/CR-3421 BIBLIOGRAPHIC DATA SHEET SAND 83-1629
- 2. (Leere ete,41 4 TITLE AND SUBTITLE (Aart Votume ho, et appreiptiew/
- Cost / Benefit Analysis of Adding a Feed and Bleed Capability to Combustion Engineering Pressurized a RECIPIENT'S ACCESSICW NO Water Reactors G. Skala 5. DATE REPORT COMPLE TED
- 7. AuTHORist D. R. Gallup E. Gahan uoNTN lveaa August 1983 R. rhordack DATE REPORT ISSUED 9 PE RF ORMiNG ORGANIZATION N AME AND MAILING ADDRE SS tinclude I.a Codel uoNTw lvtan August 1983 Sandia National Laboratories s (te.<e u.4.)
Albuquerque, New Mexico 87185 8 (Leave Nanki
- 12. SPONSORING ORGAN 12 ATION N AYE AND MAILING ADORESS (raclude 10 Codel 10 PROJECT /T ASK/ WORK UNIT NO Division of Safety Technology Of fice of Nuclear Reactor Regulation ,i ,,N No.
U.S. Nuclear Regulatory Commission Washington, DC 20555 NRC FIN No. A1309 Pt RIOD COvf #t D (f actupre aars!
13 TYPE OF REPORT Technical Report 14 fleeve osaael 15 SUPPLEMENT ARY NOTE S
- 16. ABSTH ACT 000 words or Jess)
This report presents the results of a cost / benefit analysis for the addition of a feed and blood capability to the San Onofre Nuclear Generating Station, Unit 2, (SONGS 2) . Two cases of feed 4 and bleed capability were investigated: 1) adding power operated relief valves (PORVs) to the pressurizer for depressurization and using the present high pressure safety injection (HPSI) system for reactor coolant system (RCS) inventory make-up and 2) adding an independent single train feed and bleed system. For the first case, it is estimated that the core melt frequency would be incrementally reduced by 4.0E-6 per year, a factor of 1.3, at a cost of $2.5 M to $4.3 M depending on when the equipment is installed. For the second case, it is estimated that the core melt frequency would be incremontally reduced by 1.2E-5 per year, i a factor of 3, at a cost of $7.0 M to $10.3 M. ,
i
- 17. KE Y WORDS AND DOCUME NT AN ALYSIS 17a DESCRIPTORS PORV
! Power Operated Relief Valve Feed and Bleed 3 Feedwater Reliability t
- 170 IDENTIFIERS OPEN ENDE D TERMS 19 SE CURITY CLASS (Tees reporrl 21 NO OF PAGES 18 AVAILA8tt.8TY STATEMENT Unelassified
- 2 PR#CE Unlimited 20 g,gT,Y CgSgr^d*# e s NnC FORM 335 Itisu
,C Org. Bldg. Name Rec'd by Org. Bldg, Name Rec'd by 120555078877 1-LAN US NRC ADM-DIV 0F TIDC POLICY C PUB-MGT BR-POR NUREG h-501 DC 20555 WASHINGTON.
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